Accidents in Nuclear Facilities

Accidents 2in nuclear facilities

74 Rapport scientifique et technique 2008 - IRSN 2 Accidents in nuclear facilities...... 76

2.1 Influence of enclosure on a pool fire in a mechanically ventilated compartment ...... 78

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2.2 Characterizing combustion of a liquid material ...... 89

2.3 S ummary of FPT2 TEST results from the PHÉBUS-FP program ..... 91 2.4 mephistA: a thermodynamic database on nuclear fuel developed for and applied to nuclear safety ...... 102.

newsflashnewsflashnewsflashnewsflashnewsflashnews 2.5 protecting NUCLEAR FACILITIES AGAINST AIRCRAFT CRASHES: impact studies on deformable missiles conducted by VTT in Finland ..114

2.6 sARNET: major achievements and outlook ...... 116

newsflashnewsflashnewsflashnewsflashnewsflashnews 2.7 study on iodine behavior in the reactor containment in an accident situation: first results from the EPICUR program ...... 129. 2.8 ASSESSING THE CONSEQUENCES OF MAJOR ACCIDENTS IN THE PRESENCE OF AIR: recent advances...... 131

2.9 KEY eventS and dates ...... 139.

IRSN - 2008 Scientific and Technical Report 75 ACCIDENTS in nuclear facilities

Michel Schwarz Major Accident Prevention Division

uclear facilities are complex and constantly changing, question is examined by IRSN from two complementary aspects: whether they involve electrical power generating reactors, conducting full-scale tests and developing calculation tools. Nexperimental reactors, or fuel cycle laboratories. To assess The second article illustrates basic research conducted with other measures taken by facility operators to prevent accidents or limit laboratories, in this case Ineris, to advance towards a more precise their consequences requires specialized skills and in-depth knowl- understanding of mechanisms used to calculate the heat release edge of all the phenomena involved. rate of a fire.

That is why the Institute conducts research, particularly experimen- The probability of a major accident leading to core meltdown tal research, and develops computer codes, usually through scien- is very small, since operators take the necessary measures to avoid tific cooperation with other national or international research this situation. Nonetheless, any radioactive release that could laboratories, universities, or foreign organizations providing techni- occur in this case is studied in research work conducted throughout cal support to nuclear safety authorities. the world, especially in France. Certain studies aim to validate the ASTEC numerical simulation computer code, developed in close The following articles take a brief look at the areas in which the cooperation with GRS and now a reference in Europe. Five articles Institute has invested considerable research resources. illustrate the important work conducted by the Institute in this area. An outbreak of fire in a nuclear facility, as in many other facilities, The first article presents an overview of results from the FPT2 test is a serious event that must be brought under control quickly before conducted in 2004 in CEA's PHÉBUS test reactor. This test is referred equipment items important for facility safety or for containment to as an integral test, since it represents all phenomena occurring of radioactive substances are no longer operational. For this purpose, during a water-reactor core meltdown on a reduced scale. The the Institute conducts full-scale experiments to validate the lessons learned are extremely important and have contributed advanced numerical simulation tools that it develops and uses to significantly to advancing knowledge on this subject. assess the consequences of a fire. Two articles have been dedi- The second article takes a glance at studies conducted at IRSN cated to this topic. to collect data required to reliably predict the progression of core The first presents an overview of experimental studies carried out meltdown in a reactor, which requires accurate knowledge of basic by IRSN in the last twenty years on enclosed fires in mechanically thermodynamic data on reactor components and the systems ventilated compartments, characteristic of nuclear facilities. This featuring these components.

76 2008 Scientific and Technical Report - IRSN The third article describes the success story of the European network of excellence on core meltdown accidents that the Institute has supervised for the last four years, which set out to coordinate all research work conducted in this area across Europe. The fourth article illustrates research conducted on iodine behavior by IRSN using an irradiator installed at Cadarache. Iodine-131 can present a considerable health risk in the event of accidental release. The Institute has conducted many studies and participates in several international programs relevant to this topic. The fifth article takes inventory of knowledge developed on the physicochemical behavior of ruthenium, a fission product that may lead to significant risks if released in large quantities. Physical chemistry plays a major role in determining the chemical state of a radioactive substance after it has been released by fuel in a major accident situation. The chemical state determines volatility, or the propensity of the substance to reach the containment in a gas form and then escape to the environment if leakage is present. The reactor containment provides the last barrier capable of preventing release in the event of failure of the two previous bar- riers, i.e. the metal cladding on fuel assemblies and the reactor coolant system. The last article illustrates work carried out by the Institute to assess vulnerability to impact on civil engineering structures.

IRSN - 2008 Scientific and Technical Report 77 2.1 Influence of enclosure on a pool fire in a mechanically ventilated compartment

Stéphane MELIS Fire Research and Development of Uncertainty and Simulation Methods Laboratory Laurent AUDOUIN, Hugues PRETREL Fire Experimentation Laboratory

For over twenty years, IRSN has been closely investigating fire development in the enclosed, mechanically ventilated spaces characteristic of nuclear facilities. For fuel cycle facilities, these studies aim to provide a more accurate assessment of conditions that could jeopardize containment of radioactive materials should a fire occur. There are different ways in which fire can affect containment: by suspending radioactive materials in the air inside the building; by disturbing flow in the ventilation system that normally creates a vacuum in rooms so that air circulates from the less contaminated areas to the highly contaminated areas; by damaging the HEPA filters (heated by warm gases, clogged with soot) that constitute the last filtration level before release to the environment; or by damaging the physical containment items (fire doors, fire dampers) that establish fire compartments inside the facility. To correctly simulate these phenomena, it is crucial to establish a model that accurately represents the heat released by the fire and its duration. Aside from size effects, one of the most important points is the close interaction between the development of fire in an under-ventilated environment and the air-flow behavior of the ventilation system. For several years now, this question has been approached by IRSN through two complementary lines of research: conducting full-scale tests, such as those carried out through the FLIP program (in cooperation with Areva) and the Prisme program (supported by the OECD), and the development of calculation tools such as ISIS (3D software) and the SYLVIA code, which covers the entire facility, including the ventilation system.

The heat released by fuel can usually be determined in a relatively release rate in the data set provides an acceptable agreement direct manner in well-ventilated conditions using small-, medium-, between calculations and experiment results for the main param- or large-scale calorimeters(1). In most codes designed to study eters that characterize the consequences of a fire, such as the compartment fires, entering an experimentally measured heat temperature and pressure of gases in the compartment. In actual fire situations inside a facility, however, knowing the heat release (1) The calorimeter is designed to estimate the combustion heat release rate by rate of the relevant fire type in an open atmosphere is not enough simultaneously measuring the suction flow rate of the hood and the concentrations of the various combustion products and oxygen. to predict the course it will take in an enclosed space, due to the

78 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.1

significant interaction between combustion and ventilation in the liquid pool: the surface pyrolysis rate** is related to incident room where the fire is developing. Confinement significantly fluxes in the gas phase ϕcond, ϕrad, ϕconv (by conduction, radiation, changes certain key physicochemical parameters surrounding the and convection, respectively) and to conduction/convection flux fire, such as the oxygen rate, the temperature of walls and gases towards the center of the fuel liquid ϕliq by vaporization heat and, in general, heat flux on the fuel surface. In return, these param- Lvap: eters affect both the fuel pyrolysis rate and the combustion heat, .Lvap = ϕcond + ϕrad + ϕconv - ϕliq used to determine the heat release rate(2). This feedback effect gas phase liquid phase complicates attempts to model these phenomena. Furthermore, by influencing the pyrolysis rate, fire confinement also has an effect Figure 1 illustrates the main phenomena that affect this energy on fire duration, an important parameter, since safety-related items balance as well the related modeling issues. For fuel arranged such as fire doors are designed to withstand a thermal hazard for horizontally, beyond a certain size, flame radiation becomes much a limited time period. greater than exchanges through convection and conduction in the In the nuclear industry, compartments are generally hermetically gas phase. The radiated flux is closely related to soot formation; separated from each other, but remain interconnected by a venti- precursors of soot are produced above a certain temperature (around lation network. The mechanical ventilation system provides a 1,350 K) and the aggregates that constitute the complete soot cascade of vacuum zones designed to prevent any accidental leak- formation contribute primarily to the flame radiation properties. age of radioactive substances to the environment. A vacuum is Therefore as soot formation decreases, so does the fuel pyrolysis maintained in the rooms containing radioactive materials, with the rate. Detailed modeling of soot formation and oxidation must take exhaust air passing through HEPA filters. This particular configura- into account an extremely complex chemical process, which is often tion plays an important role in a fire situation, since it quite often slow compared to the time characteristic of flows in the flame, and leads to "under-ventilated" fires, i.e. fires that develop in an oxygen- must also provide an accurate representation of the temperature starved atmosphere, since the ventilation system supplies less fields in the flame, while calculating the associated aggregate oxygen than what would be consumed by the fire in the open air. formation and radiation; this is why the phenomena that influence It is precisely in under-ventilated combustion conditions that the the pyrolysis rate, such as soot formation, are often modeled physicochemical parameters in the surrounding fire environment analytically or empirically. have the most influence on combustion. Size effect For fires measuring more than 10 cm, an increase in size entails an Fuel pyrolysis increase in the zone containing soot. The flame is then more opaque and radiates more towards the fire. This also explains why large fires For over twenty-five years, numerous researchers have studied the are more sensitive to radiation from the ambient environment. influence of confinement on the heat released by a pool fire. According to Hottel and Babrauskas [Babrauskas, 1983], the size of Experiments conducted by Takeda, Fleischman and more recently the soot formation depends linearly on the size of the fire (D), and by Quintiere [Utiskul et al., 2005] focus on small-sized compart- a correlation links the pyrolysis rate per unit of surface area bv 3 ments (a maximum of 1 m ), with natural ventilation, usually to that of very large fires ∞ through a coefficient kß, which takes consisting of vertical openings that simulate an open door. This into account both the propensity of the fuel to produce soot and configuration was used to identify various combustion conditions, the optical properties of the soot aggregates. depending on the size of the openings, and whether they were non-oscillating, steadily oscillating, or unsteadily oscillating. Pyrolysis rates up to seven times greater than those observed in Figure 2 shows the good agreement between this correlation and the open air were reported, the influence of the additional radiation measurements taken on TPH(3) under the SATURNE calorimeter due to confinement perhaps explaining these observations. On a hood at IRSN, for different combustion tank sizes [Pretrel, 2007]. larger scale [Peatross and Beyler, 1997] or in configurations where the oxygen rate was controlled using nitrogen dilution [Tewarson . . et al., 1981], a linear decrease in the pyrolysis rate in parallel with (2) The fire heat release rate is given by the equation Q = m × ∆Hc , where ∆Hc is combustion heat, i.e. the heat released by the reaction of one kilogram of . the oxygen concentration was, however, observed. fuel, and m is the pyrolysis rate, more commonly expressed as the product of . the fire surface area and the surface pyrolysis rate** m ". Fuel pyrolysis is the result of fuel vaporization, an endothermic (3) Hydrogenated tetrapropylene (TPH) is a dodecane (C12H26) used as a solvent in process. It is based on the heat flux balance at the surface of the the nuclear industry.

IRSN - 2008 Scientific and Technical Report 79 2.1 e Thermal plum Thermal

Increasing the fire size Air entrainment causes the soot generating in the flame Soot oxidation zone area to grow larger. Flame emissivity intensifies, driving Stoichiometric zone heat radiation towards the Soot generating zone fuel.

The flame temperature For horizontal fuels sized depends on the ambient flame Intermittent greater than 10-20 cm, the oxygen concentration. main mechanism that transfers the heat of the flame towards the fuel is radiation. At the extinction boundary (O2 = 12 vol %), the temperature is too low to Conduction inside the tank generate soot. The flame (at the beginning of the fire) turns bluish, indicating that and the energy required for radiation exchanges are low. fuel pyrolysis complete the energy balance for the flame Steady combustion tank. Radiation Conduction Convection Fuel

Conduction Pyrolysis

Figure 1 Ideal representation of the flame and the main mechanisms that influence the fuel pyrolysis rate.

Effect of decreasing oxygen in the ambient environment Transient effects on fire ignition The technical literature indicates a linear dependence between the In experiments, the fire is ignited artificially using either a radiating pyrolysis rate ( ) and the oxygen concentration in the ambient panel or a small pilot flame aimed at the fuel tank. The experiment environment (Y). itself does not begin until after this startup phase, which lasts a few minutes. When the fire ignites, two phenomena tend to gradually increase the fuel pyrolysis rate. First, the flame propagates to the This rate is expressed in terms of the corresponding rate represent- entire combustion tank at a relatively high spreading rate (a few

ing well-ventilated conditions (indicated by bv, where Ybv = 0.233 cm/s), so that the tank is quickly completely in flames. Second, the for standard conditions) with a linear dependence coefficient α fuel liquid, which is initially cold, thereby forming a heat sink, is that varies depending on the author: Peatross and Beyler [1997] gradually heated to its center through conduction by the radiated propose α = 1.1 while others [Utiskul et al., 2005] set α = 0, but heat received at the surface, which prompts vaporization. Assuming add an ambient environment temperature function to this correla- a constant incident heat flux and observing that the thermal behavior tion. This dependence is illustrated in Figure 3 for a test conducted of the liquid in the tank is similar to that displayed in a semi-infinite through the PRISME SOURCE program: at any given instant, the environment, it is possible to establish an analytical solution of the experimental measurements of the pyrolysis rate and the oxygen liquid warming conditions that expresses the increase in pyrolysis concentration near the fire corroborate the Peatross and Beyler rate as a function of time and the thermophysical properties of the correlation for this configuration. fuel, its initial temperature and the tank dimensions.

80 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.1

** Surface pyrolysis rate (g/m²/s) Normalized pyrolysis rate (–) 60 1

50 0,8

40 0,6

30 (Ignition phase) 0,4 20 (Quasi-steady state)

0,2 10

(Extinction phase) 0 0 0,10 110 10 12 14 16 18 20 Tank diameter (m) Oxygen (vol %)

S1 S3 S5 LIC2CB Test 1 Peatross et al. Utiskul et al. 48 [1-exp(-1.8 D)] 52 [1-exp(-1.6 D)]

Figure 2 How size affects pyrolysis rate for different tanks Figure 3 How decreasing the oxygen concentration affects the containing TPH and the corresponding Babrauskas laws. pyrolysis rate in Test 1 (0.4 m2 of TPH in the DIVA facility, PRISME SOURCE program), with the corresponding correlations.

Figure 4 compares this model, currently being validated at IRSN, Pyrolysis rate (kg/s) with the test results obtained using the SATURNE hood for different x 10 -2 tank sizes in the context of the PRISME SOURCE research 2 1.8 program. 1.6 1.4 Fire extinction due to oxygen starvation 1.2 As the oxygen concentration in the ambient environment decreas- 1 0.8 es, extinction phenomena appear, first limited to certain areas 0.6 within the flame, then leading to complete extinction of the fire. 0.4 Two approaches are currently proposed to model these phenom- 0.2 ena. The first (see [Beyler, 2005] for a bibliography review) is based 0 on the fact that the flame temperature decreases as the oxygen - 0.1 00.1 0.20.3 0.410.50.6 0.70.8 0.911.1 1.21.3 1.41.5 1.61.7 .8 1.9 2 x 10 3 concentration drops due to dilution of the comburant, until the Time (s) flame is extinguished. The second approach [Delichatsios, 2007] SYLVIA S5 (0.1 m2) Exp S5 (0.1 m2) models heat exchange between the flame and the fuel, and infers SYLVIA S3 (0.4 m2) Exp S3 (0.4 m2) that there is a critical mass transfer number that determines flame SYLVIA S2 (0.2 m2) Exp S2 (0.2 m2) extinction. At present, it is difficult to choose between these two approaches. An engineer's solution would consist of applying an Figure 4 Modeling pyrolysis rate increase on ignition for TPH tanks of three different sizes burning in the open air. extinction criteria based on a threshold value of the surface pyrolysis rate (about 10 g/m2/s), deduced from a large set of liquid pool fire test results. It would then be necessary to establish the theoretical principles that justify this value.

IRSN - 2008 Scientific and Technical Report 81 2.1

South wall Dilution

7.40 m NE thermocouple fixture NW thermocouple fixture Exhaust North door Exhaust Filters Fan East wall West wall Tank Pool fire Z 400 m3 X Y Inlet 6 m North wall

9 m Air inlet

Figure 5 PLUTON facility (400 m3).

Ventilation system 6 m

Admission Exhaust East

0.75 m 0.4 m

3.35 m 3.95 m 5 m

0.4 m

West

Fire - side view Fire – view from above

Figure 6 DIVA facility (120 m3).

Global experiments to validate models The test results provided very interesting data on how confinement affects combustion, frequently revealing a relatively long quasi- Studying pool fires in an enclosed environment led to several tests steady state. Certain experiments were selected to validate the conducted at IRSN. Their specificity resides, first, in the size of the models, and their main characteristics are listed in Table 1. compartments used in the test setup, and second, in the use of a All the experiments were instrumented with type K thermocouples mechanical ventilation system. Two facilities located at the positioned on the vertical measuring axes to provide a precise Cadarache site were used: PLUTON (Figure 5) a 400 m3 concrete map of the gas temperatures inside the compartment. Oxygen enclosure 7.5 m high, and DIVA (Figure 6) consisting of several concentration was measured using Servomex 4100 analyzers at concrete enclosures measuring 120 m3 in volume, with walls 4 m three points in the compartment where the fire developed: one high. The PLUTON ventilation system was adjusted so that air-flow near the fire, 35 cm above the floor, another located 80 cm resistance was greater at the inlet than at the outlet. The adjust- above the floor, and the last one placed 70 cm below the ceiling. ment made on the DIVA facility was more balanced. Different An "average" oxygen concentration in the compartment was experimental conditions were studied on these facilities by chang- estimated, based on measurements from the second and third ing the ventilation rate, the type of fuel, and the size of the analyzers. Table 2 summarizes the main results from these combustion tank. experiments.

82 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.1

Initial Global equiv- Tank surface ventilation Fuel mass Air inlet Test Fuel Facility alence ratio area (m2) flow rate (kg) position (kg/s) Φ 1(a) 0.4 TPH DIVA 0.183 15.0 high 0.97 2(a) 0.4 TPH DIVA 0.337 15.7 high 0.53 3(a) 0.4 TPH DIVA 0.192 15.7 high 0.93 4(a) 0.2 TPH DIVA 0.187 7.2 high 0.39 5(a) 0.4 TPH DIVA 0.187 16.0 low 0.95 6(a) 0.4 TPH DIVA 0.067 15.8 low 2.66 7(b) 1 TBP/TPH(c) PLUTON 0.247 41.2 high 1.73 8(b) 0.4 TBP/TPH PLUTON 0.4 16.7 low 0.37 9(a) 0.56 Oil(d) DIVA 0.12 20.0 high 0.91

Table 1 Test characteristics. (a) [Pretrel and Querre, 2005a]. (b) [Audouin and Tourniaire, 2000] ; [Pretrel and Such, 2005b]. (c) Tank placed against a wall. (d) Liquid preheated to 250 °C before ignition.

Average oxygen con- Average oxygen con- Pyrolysis rate in the Pyrolysis rate in the centration in the fire Average temperature centration near the Test open air quasi-steady state compartment during of gases in the quasi- fire during the quasi- (g/m2/s) (g/m2/s) the quasi-steady state steady state (K) steady state (vol %) (vol %)

1 34.7 10.0 14.9 14.2 435 2 34.7 15.0 17.2 15.5 470 3 34.7 10.5 14 14.2 435 4 28.7 14.5 18.0 16.3 400 5 34.7 16.0 11.7 16.1 475 6 34.7 10.5 13.2 13.5 435 7 38.2 8.3 14.9 n/a 448 8 31.8 20.0 16.6 18.5 399 9 28.6 9.3 14.3 14.8 428

Table 2 Main results.

’’bv ’’ All tests showed a temperature rise that was limited compared 6 to the temperatures leading to flashover(4), while the oxygen concentration often decreased significantly. This made it possible 5 to study the influence of the oxygen concentration on the pyrol- 4 ysis rate, while ignoring the temperature effect. 3

Two experiments (Figures 6 and 7) led to early extinction of the 2 fire due to oxygen starvation, while the other tests displayed a 1 phase in a quasi-steady state characterized by a pyrolysis rate much lower than that observed in the open air for the same type 0 0 0.5 1 1.5 2 2.5 of tank. Global equivalence ratio Φ

Theory TPH, high injection ** Oil, high injection TBP/TPH, high injection TPH, low injection TBP/TPH, low injection (4) Flashover occurs when all combustible materials in the room ignite simultaneously. It is usually assumed that a temperature of 875 K (approx. 600°C) in the upper Figure 7 Reciprocal of pyrolysis rate in quasi-steady state as a part of the room will lead to flashover. function of the global equivalent ratio.

IRSN - 2008 Scientific and Technical Report 83 2.1

’’ Mass flow (kg/s) ’’bv x 10 -2 1 1.4

1.2 0.8

1 0.6 0.8

0.4 0.6

0.2 0.4

0.2 0 12 13 14 15 16 17 18 19 20 21 22 0 O near the fire (vol %) 2 0 0.5 1 1.5 2 2.5 3 3.5 x 10 3 Beyler (3) Time (s) Oil, high injection TPH, high injection Code xp TPH, low injection TBP/TPH, low injection

Figure 8 Peatross and Beyler correlation (α = 1.1). Figure 9 Pyrolysis rate (Test 1).

Validation of the "Well-mixed Reactor the cascade of vacuum zones operational in the normal state, such

approach" that q /q0 = 0.9 is a satisfactory approximation; (5) the global equivalence ratio where Af is the The "Well-mixed Reactor" approach consists of considering the fire fire surface area and s the oxygen mass consumed by the fuel mass compartment as a reactor where properties are homogeneous. burned in stoichiometry. This ratio expresses the relationship Simple mass and energy balances are therefore enough to determine between the oxygen required for burning and the oxygen supplied the general characteristics of the gases and walls in the compart- in the normally-ventilated state. ment. This approach is used to characterize fires after flashover, and also to conduct dimensional studies aimed at establishing the Figure 7 suggests two types of combustion behavior in the quasi- primary parameters for modeling. Applying this approach to the steady state: the tests where the air inlet is in the upper part of the case of a fire confined in a ventilated compartment leads to an compartment reveal a remarkable degree of agreement with the equation that gives the pyrolysis rate in a quasi-steady state [Melis well-mixed reactor theory, whereas a significant difference is and Audouin, 2008]. By considering that the admission and exhaust observed when the inlet is located near the floor: branches follow a Bernoulli law, the conservation of mass, oxygen, when the air inlet is near the ceiling, fresh air is convected towards and energy in the compartment makes it possible to determine the the floor of the compartment, mixing the gases so that gas layering oxygen fraction during the quasi-steady state, and thus the pyrol- is attenuated; ysis rate. in opposition, enclosed fires where the air inlet is near the floor of the compartment behave similarly to fires naturally ventilated where by an open door, where significant thermal layering of gases is P0 - Pdownstream P - P observed. The fire and the lower part of the compartment are upstream downstream immersed in a practically non-vitiated environment favorable to This equation establishes a relationship between the fuel pyrolysis combustion and supplied with fresh air, while the combustion rate during the quasi-steady phase ( ) and the pyrolysis rate in a products accumulate in the upper part of the compartment. well-ventilated environment ( ). It displays three terms: The difference between the mean oxygen concentration and the the oxygen concentration dependence coefficient α; oxygen concentration near the fire (Table 2) confirms this analysis.

the ratio q /q0 representing the ventilation rate in the quasi-steady state over the nominal ventilation rate of the facility. This ratio (5) The Global Equivalence Ratio (GER) was introduced by Tewarson and characteriz- depends to a slight degree on the compartment temperature and es under-ventilation of a fire in comparison to stoichiometry.

84 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.1

Fraction (–) mgl Mass flow (kg/s) x 10 -2 0.20 1.4 0.18 1.2 0.16

0.14 1

0.12 0.8 0.10

0.08 0.6

0.06 0.4 0.04 0.2 0.02

0 0 0 0.5 1 1.5 2 2.5 3 3.5 4 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.61.8 22.2 2.42.6 2.83 x 10 3 x 10 3 Time (s) Time (s) Code CO2 xp CO2

Code CO2 xp CO2 Code xp

Figure 10 Oxygen and CO2 concentrations (Test 1). Figure 11 Pyrolysis rate (Test 2).

Yet in both cases, the Peatross and Beyler correlation (α = 1.1) can The ventilation system is modeled as an network of branches and be applied satisfactorily (Figure 8). nodes. The branches represent the ventilation ducts, fans, filters, etc. The main unknown is the gas flow rate, the relevant equation The "Well-mixed Reactor" approach therefore cannot be used to being a general form of the Bernoulli equation that takes into predict fire behavior for all configurations. It nonetheless reveals account the inertia of momentum. Nodes represent either the the global equivalence ratio as a dimensionless value characteristic compartments or the connection points between branches. The of the fire: this number is used to theoretically classify fire experi- conservation of mass for each specie and the conservation of ments and scenarios and plays an important role in describing the energy make it possible to determine the thermodynamic proper- qualification limits of computer codes. ties of the gas in the branches and nodes. Walls are meshed in two dimensions, but heat conduction is only resolved for thickness. A gradual representation where meshing is Validation of the Lumped-Parameter tighter near the surface is used to take into account significant Approach (SYLVIA code) temperature gradients in the walls. Heat transfer takes place through natural convection in compartments and through forced convention Description of the Lumped-Parameter Approach in branches. Heat transfer coefficients are obtained using conven- "The Well-mixed Reactor" approach shows that it is essential to tional correlations from the technical literature. Heat exchanges correctly estimate the oxygen available for combustion in order to through radiation involve the flame, walls, and gases. The model correctly model the fire heat release rate. Models based on the used is a hot-spot model where the fire is represented as a hot spot Lumped-Parameter Approach take into account thermal gas layer- emitting radiated heat that is a constant fraction of the fire heat ing in an enclosed fire in a simplified manner. The space inside the release rate. Absorption and emission by gases take into account compartment is divided into two zones, the height of the interface carbon dioxide, steam, and soot content. between these zones changing over time. In each zone, the ther- One of the unique features of SYLVIA is its ability to calculate the modynamic properties (gas pressure and temperature, concentration pyrolysis rate, which must be provided as input for most computer of the various species) are assumed to be uniform. The SYLVIA code, codes used in fire modeling. developed by IRSN, was inspired by this simplified compartment The various phenomena that affect the pyrolysis rate are processed model that nonetheless gives a detailed description of the entire using correlations: the size effect is calculated using the Babrauskas ventilation system. correlation; the impact of decreasing oxygen concentration is processed by the Peatross and Beyler correlation, where the oxygen

IRSN - 2008 Scientific and Technical Report 85 2.1

Fraction (–) mgl Mass flow (kg/s) x 10 -2 0.20 1.4 0.18 1.2 0.16

0.14 1

0.12 0.8 0.10

0.08 0.6

0.06 0.4 0.04 0.2 0.02

0 0 0 0.5 1 1.5 22.5 33.5 0 0.5 1 1.5 22.5 33.5 4 x 10 3 x 10 3 Time (s) Time (s) Code CO2 xp CO2

Code CO2 xp CO2 Code xp

Figure 12 Oxygen and CO2 concentrations (Test 2). Figure 13 Pyrolysis rate (Test 5).

concentration used is interpolated from the concentrations in the An example of this configuration is Test 5, illustrated in Figures 13 upper zone and lower zone, according to the height of the interface and 14. The slight deviation with respect to the Peatross and Beyler between the two zones; transient effects at fire ignition are calcu- correlation (Figure 7) is not enough to explain the gap between lated based on the thermophysical properties of the liquid fuel; calculation results and experimental results. In this case, it is the lastly, fire extinction occurs either after all the fuel has been consumed, chemical process in the reaction that varied significantly from or when the surface pyrolysis rate** goes below a critical value. well-ventilated conditions: when the upper part of the flame was surrounded by a vitiated atmosphere, combustion became less The combustion reaction and combustion heat are considered as exothermic, consumed less oxygen, and produced more carbon constant, thereby establishing a linear relationship between the monoxide and soot. This phenomenon (not modeled) was revealed

pyrolysis rate, the formation of combustion products, oxygen by observing the CO/CO2 ratio, which was ten times greater in Test consumption, and heat release. 5 than in Test 1 (2% and 0.2% respectively), the only difference between these two tests being the position of the air inlet. Validation of the SYLVIA code Systematic comparison between predictions obtained using the In more general terms, confinement can have two effects with SYLVIA code and results from full-scale tests conducted at IRSN opposite consequences, depending on the value of the global led to broad validation of this computer application. Figures 9 to equivalence ratio: 14 illustrate the enclosed pool-fire dynamics. Two examples repre- for slightly under-ventilated fires, the decrease in the pyrolysis sentative of tests where the air inlet duct was located in the upper rate entails an increase in the fire duration in comparison to the part of the compartment are given for Tests 1 and 2 (Figures 9 to same fire in the open air, thereby aggravating the thermal impact 12). The pyrolysis rate as well as the oxygen and carbon dioxide of the fire; concentrations were calculated quite satisfactorily in the quasi- in the opposite case, when under-ventilation is significant, i.e. steady state, although the representation of pyrolysis behavior at when the global equivalence ratio is high, fuel pyrolysis is not suf- fire ignition was rather approximate. Agreement was not as good ficient to sustain combustion and the fire is extinguished before all when the air inlet duct was located in the lower part of the com- the fuel has been consumed, thus attenuating the thermal impact partment, the lumped-parameter approach providing better results of the fire. Tests 6 and 7 are representative of this early extinction than the "Well-mixed Reactor" approach. effect.

86 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.1

Fraction (–) mgl Calculated fire duration (s) 5,000 0.20

0.18 4,000 0.16

0.14 3,000 0.12

0.10 2,000 0.08

0.06 1,000 0.04

0.02 0 0 500 1,000 1,50022,0003,500 3,0004,500 4,0005,500 ,000 0 Experimental fire duration (s) 0 0.5 1 1.5 22.5 33.5 4 x 10 3 Time (s) id TPH, high injection Oil, high** injection TBP/TPH, low injection Code CO xp CO 2 2 TPH, low injection TBP/TPH, high injection Code CO2 xp CO2

Figure 14 Oxygen and CO2 concentrations (Test 5). Figure 15 Comparison of calculated and experimental fire duration periods from all tests conducted.

As shown in Figure 15, these two effects were correctly predicted accurately quantify the effect of vitiation on the pyrolysis rate using a simple lumped-parameter approach coupled with the of liquid fuels. Work in this field must be pursued to study the Peatross and Beyler correlation. The model nonetheless greatly validity of these models for carbonaceous fuels and vertical overestimates fire duration for Test 5. fires.

Furthermore, vitiation of the environment near the fire may Conclusion also entail extinction of the fire. An extinction criteria based on the fuel surface pyrolysis rate** would make it possible to Modeling an enclosed fire requires, above all, correctly predict- obtain a satisfactory model of this phenomenon. ing the heat released through combustion, therefore the fuel pyrolysis rate. For well-ventilated environments, the technical literature proposes correlations that take into account the various phenomena that affect pyrolysis. The size effect and transient aspects of fire ignition were studied on liquid fuels of interest with regards to nuclear safety in fire studies conducted using the SATURNE calorimeter hood. For slightly ventilated environments, several experiments that studied the effect of fire confinement were conducted at IRSN using the PLUTON and DIVA facilities, consisting of closed compartments connected to a ventilation system. The main effect of enclosure can be expressed as a linear decrease in the pyrolysis rate occurring simultaneously with the decrease in the oxygen concentration. The technical literature, however, reports different values for the linearity coefficient, the one proposed by Peatross and Beyler being in closest agreement with experiments conducted at IRSN. It was therefore possible to more

IRSN - 2008 Scientific and Technical Report 87 2.1

References L. Audouin, B. Tourniaire (2000). Incident heat fluxes from pool fires close to a wall in a well-confined compartment, 8th int. conference on nuclear engineering. V. Babrauskas (1983). Estimating large pool fire burning rate, Fire technology, 19 : 251-261. V. Babrauskas (2002). Heat Release Rates, The SFPE Handbook of Fire Protection Engineering (3rd ed), DiNenno P.J. (ed.), National Fire Protection Association, Quincy, MA 02269, 2002, p. 3/1. C.-G. Beyler (2005). A brief history of the prediction of flame extinction based upon flame temperature, Fire Safety Journal 29 : 425-427. M.-A. Delichatsios, M.-M. Delichatsios (1997). Fire Safety Science-Proceedings of the Fifth International Symposium, International Association for Fire Safety Science, 1997, p. 153-164. M.-A. Delichatsios (2007). Surface extinction of flame on solids: some interesting results, Proceedings of the 31st Symposium International on Combustion, The Combustion Institute, 2007, p. 2 749-2 756. S. Melis, L. Audouin (2008). Effects of vitiation on the heat release rate in mechanically-ventilated compartment fires, Fire Safety Science-Proceedings of the Ninth International Symposium, International Association for Fire Safety Science, 2008. M.-J. Peatross, C.-L. Beyler (1997). Ventilation Effects on Compartment Fire Characterization, Fire Safety Science-Proceedings of the Fifth International Symposium, International Association for Fire Safety Science, 1997, p. 403-414, doi:10.3801/IAFSS.FSS.5-403. H. Pretrel, P. Querre (2005a). Experimental Study of Burning Rate Behaviour in Confined and Ventilated Fire Compartments, Fire Safety Science-Proceedings of the Eighth International Symposium, International Association for Fire Safety Science, 2005, p. 1 217-1 229, doi:10.3801/IAFSS.FSS.8-1 217. H. Pretrel, J.-M. Such (2005b). Effect of ventilation procedures on the behaviour of a fire compartment scenario, Nuclear engineering and design 23 : 2 155-2 169, doi:10.1016/j.nucengdes.2005.03.003. H. Pretrel (2007). Phenomenological study of large scale pool fire behaviour in free atmosphere, 11th International Conference on Fire Science and Engineering, Interflam 2007. J.-G. Quintiere, A.-S. Rangwala (2004). A theory for flame extinction based on flame temperature, Fire and Materials 28:387-402, doi:10.1002/fam.835. A. Tewarson, J.-L. Lee, R.-F. Pion (1981). The Influence of Oxygen Concentration on Fuel Parameters for Fire Modeling, Proceedings of the Eighteenth Symposium (International) on Combustion, Combustion Institute, 1981, p. 563-570, doi:10.1016/S0082-0784(81)80061-6. Y. Utiskul, J.-G. Quintiere, A.-S. Rangwala, B.-A. Ringwelski, K. Wakatsuki, T. Naruse (2005). Compartment Fire Phenomena under Limited Ventilation, Fire Safety Journal 40 : 367-390, doi:10.1016/j.firesaf.2005.02.002.

88 2008 Scientific and Technical Report - IRSN newsflashnewsflashnewsflashnewsflashnewsflashnewsflash

Characterizing combustion 2.2 of a liquid material

Laurent AUDOUIN Fire Experimentation Laboratory Mass flow per unit area (g/m2/s) Stéphane DUPLANTIER 60 Accident Risk Division (Ineris) rr rr 50 m’ = m’ Te wa r son (47% O2)

40 Babrauskas’ correlation 30 for TPH Experiment data 20

0 0 0.5 1 1.5 232.5 Liquid pool diameter (m)

Figure 1 Fuel mass flow per unit area as a function of tank diameter.

As a fire burns, the heat released by vapor- unit area as a function of the liquid pool ized fuel combustion determines how the diameter (see example in Figure 1). The fire will develop, along with fire character- database containing mass flow per unit area istics such as the flame height above the for combustible materials serves to run fire fire, the energy radiated from the flame to scenario simulations using computer codes. the immediate environment, the gas temperature and velocity in the flame, and the Since it is easier, faster, and less costly to smoke plume. In practice, this thermal power run tests on a small scale rather than a large can be estimated by multiplying the fuel one, in the last few years researchers have mass flow per unit area by the fire surface been developing an experimental protocol area and the combustion heat, a thermody- to estimate the mass flow per unit area of a namic property of the material. The mass large-sized fire (with a diameter greater than flow per unit area of a liquid fuel depends one meter) based only on small-scale tests. on the size of the tank containing the fuel, Ineris(1) and IRSN, working in partnership as illustrated in Figure 1 for hydrogenated on this research subject, have contributed tetrapropylene (TPH), a liquid hydrocarbon to achieving the first step by proposing a used in the nuclear fuel recycling process. method that determines the asymptotic The mass flow per unit area of a fuel is mea- value of the mass flow per unit area for sured using a calorimeter cone for tank liquid fuels on a large scale (noted ) based diameters ranging from roughly 10 cm on tests conducted in a Tewarson calorim- (Figure 2a) to several meters (Figure 2b). eter. The method consists of measuring the Experimental results obtained in this way mass flow per unit area on a material burn- can be used to determine the parameters ing in an over-ventilated environment (i.e. (1) Ineris is the French national institute for the study of industrial environments and risks. of a correlation that gives the mass flow per where the volume fraction of oxygen is

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a b

Figure 2 a) Small-scale Tewarson calorimeter (Ineris) (tank diameter approx. 10 cm). b) Large- scale calorimeter (IRSN) (tank diameter approx. 1 m).

greater than 21%), as proposed in the FRST extending the scope of validity of this new technique(2). empirical approach to other liquid fuels and developing a model of the physicochemical The study conducted with Ineris showed phenomena involved. More long-term that the level of over-ventilation required research goals aim to verify whether this to reach the asymptotic value of the approach can be extended to solid fuels, where mass flow per unit area is correlated to the thermal degradation is more complex (pyrol-

adimensional parameter cs. This parameter ysis with surface oxidation, carbonaceous represents the relationship between gasifi- residues, etc.). cation heat and combustion heat, which are thermodynamic properties of the material. The first results obtained through this In practice, the value of a material with scientific partnership between Ineris and

an cs equal to 0.018 (such as TPH, for exam- IRSN appeared in a joint publication in the ple) is obtained by burning a sample of this 11th Proceedings of the International material in a Tewarson chamber in an atmo- Interflam Conference held in London in sphere containing 47% oxygen by volume September 2007. (Figure 1). (2) FRST: Flame Radiation Scaling Technique. Approach To date, this experimental technique for proposed in the 1980's by Tewarson, based on the empirical observation that a material burning in determining mass flow per unit area on a an over-ventilated environment manifests a mass large scale has been verified successfully on flow per unit area that is higher than in an ambient environment. nine liquid fuels. The next step consists of

90 2008 Scientific and Technical Report - IRSN 2.3 Summary of FPT2 TEST results from the PHÉBUS-FP program

Béatrice SIMONDI-TEISSEIRE Department of Studies and Experimental Research on Chemistry and Fires

Since the accident at the Three Mile Island nuclear power plant Unit 2 (TMI-2) in the US on March 28, 1979, which led to partial core meltdown and limited fission product release, several organizations around the world have reinforced their research on safety by developing a number of experimental programs. The PHÉBUS-FP experimental program, conducted on the CEA PHÉBUS reactor, was initiated by IPSN in 1988 and is one of the major international research programs dedicated to severe accidents (involving core meltdown) on water reactors. The program involved a series of integral experiments reproducing all expected physical core meltdown phenomena as realistically as possible. Experimental results from the PHÉBUS-FP program combined with results from separate effect tests(1) are essential to validating the simulation codes used in light water reactor safety analyses [Birchley et al., 2005; Clément, 2003a; Clément et al., 2006; Evrard et al., 2003; Schwarz et al., 1999; Schwarz et al., 2001], in particular the ASTEC code [Van Dorsselaere and Allelein, 2004], developed by IRSN in collaboration with GRS, as well as the ICARE/CATHARE code.

The PHÉBUS-FP program consisted of five tests successfully Experimental facility conducted from 1993 to 2004 [Clément et al., 2006]. FPT2 was the fourth test in the program, carried out from October 12 to The PHÉBUS facility is designed to create experimental conditions 16, 2000. The study focused on core meltdown under steam- representative of a core meltdown accident in a pressurized starvation conditions, where the steam contained boric acid, and water reactor [Schwarz et al., 1999; Clément et al., 2003b], in the sump was basic and evaporating, whereas in previous tests order to investigate the degradation of fuel rods and a neutron- steam was in excess and the sump was acid and non-evaporat- absorbing rod(3) up to the formation of a molten pool. It is also ing(2). used to study the release and transport of materials (fuel fission products, vapors, or aerosols from fuel bundle degradation) The experimental data collected during the test and subsequent resulting from deterioration of the primary circuit and the destructive and non-destructive test campaigns was processed and overall data consistency analyzed. Findings from the FPT2 (1) Studies conducted on a limited number of phenomena. (2) These conditions are chosen according to the accident sequence to be test results [Clément et al., 2006; March, 2008; Grégoire et al., simulated. 2008] are presented in this article. (3) Used in nuclear reactors to control reactor power.

IRSN - 2008 Scientific and Technical Report 91 2.3

PWR

7 Painted Scale = 1:5000 surface 3

Containment PHÉBUS-FP 5 3 7

2 1 4 5 Primary Break 2 circuit 4

6 6

Figure 1 1 Bundle consisting of fuel rods and the absorber rod. 2 Experimental circuit hot leg. 3 Inverted U-shaped tube simulating the steam generator. 4 Experimental circuit cold leg. 5 Gas containment volume in the vessel simulating the reactor containment. 6 Tank containing liquid to simulate the sump. 7 Painted surfaces.

containment. Researchers are particularly focused on iodine These three systems are simulated at a scale of approximately behavior, since environmental release of this radionuclide in the 1:5000 with respect to a 900 MWe pressurized water reactor days following core meltdown could have considerable radio- (Figure 1, left-hand side). logical consequences. The test facility is equipped with various instruments to measure flow rate, temperature, radiation (high-count gamma spectrom- FPT2 investigated physical phenomena occurring in the follow- etry), concentrations of hydrogen and oxygen, and to take ing plant systems (see Figure 1, right-hand side): sequential samples of circuit fluid, containment atmosphere, the reactor core, simulated by a fuel bundle containing and sump liquid. Three thermal gradient tubes (TGT), with a wall (4) 18 UO2 fuel rods previously irradiated in the BR3 reactor with temperature profile set from 700°C to 150°C, were installed on a burnup of 32 GWd/tU, two instrumented rods containing fresh the hot leg to determine the element condensation temperatures

UO2 fuel, and a neutron-absorber silver-indium-cadmium (SIC) and thus deduce their chemical speciation. Non-destructive rod. The fuel rods have Zircaloy cladding(5) and the absorber rod post-test measurements were performed in the facility to quan-

has steel cladding with a Zircaloy guide tube ( 1 ); tify the gamma emitters retained in the experimental circuit the primary circuit, represented by a hot leg where the wall and vessel samples, and to characterize fuel degradation (using

temperature is controlled to 700°C ( 2 ), and a cold leg controlled X-ray radiography, computed tomography(8), and gamma spec- to 150°C ( 4 ), interconnected by an inverted U-shaped tube four trometry to establish the γ-emitter distribution profile of the

meters high, simulating the steam generators ( 3 ); bundle). Destructive measurements were also carried out on the the containment building, simulated by a 10 m3 vessel(6) with fuel bundle [Bottomley et al., 2007] and on a selection of sam- an electropolished surface, a 120-liter tank at its base filled with

a pH 9 alkaline buffer solution simulating the reactor sump (4) Reactor operated by Belgonucléaire in Mol, Belgium.

( 6 ), a gas containment volume ( 5 ) and, in the upper part, (5) Zirconium alloy. (6) Referred to as the "vessel" in the rest of this article. condensing, cooled, painted surfaces(7) ( 7 ). The cold leg dis- (7) Referred to as "condensers" in the rest of this article. charges into the containment, simulating a break downstream (8) Cross-sectional views of the experimental cell, where each pixel is assigned a from the steam generator. specific density.

92 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.3

ples taken from the experimental circuit and the vessel to obtain to recombine hydrogen into water, thereby avoiding the risk of further information on bundle degradation and aerosol release. explosion. Most of these objectives were reached. However, since most of the flow meters associated with samplings on the hot leg did FPT2 test objectives not operate, , the release flux curves of the various elements in the hot leg could not be established. In return, on-line gamma FPT2 set out to achieve the specific objectives described spectrometry measurements taken on the vessel and cold leg below. flux curves established using post-test gamma spectrometry measurements taken on samples generally showed good consis- On the bundle tency. They provided reliable quantitative data on kinetics and The test aimed to achieve significant degradation of the fuel the element quantities recovered in the vessel. It should be noted rods and the absorber rod by melting up to 20% of the fuel, or that boron was not measured on the samples. roughly 2 kg. It also set out to observe the consequences of the release of fission products and structural materials by the bundle, in a FPT2 test scenario hydrogen-rich, low-pressure atmosphere (0.2 MPa) containing boric acid. Prior to the actual experimental phase, the bundle was "re-irradiated" in the PHÉBUS reactor for eight days to obtain In the experimental circuit a representative inventory of short-lived fission products (such The main goal was to study the transport and retention of fission as iodine-131, with a half-life of around eight days). Then the products and structural materials in the primary circuit at low bundle dried for about 37 hours and boundary conditions were pressure (0.2 MPa) in a highly reducing atmosphere(9). Data was to adjusted. During the ensuing degradation phase, which lasted be obtained on fission product chemistry (particularly on the influ- about six hours, pressure in the experimental circuit was held at ence of boric acid), and the interaction between fission products 0.2 MPa and the steam injection flow rate in the lower portion and the circuit walls at high temperature, especially in places where of the bundle was maintained at 0.5 g/s (boric acid with a boron significant temperature changes were expected (at the outlet of mass concentration of 1,000 ppm was added to the steam). the bundle and at the inlet of the steam generator tube). Reactor power was increased in a stepwise manner, leading to gradual degradation of the test bundle (Figure 2). Once the In the containment vessel degradation targets had been reached, the reactor was shut The main objective was to study fission product chemistry, down. After this transient, the fuel bundle was then cooled for especially iodine, in the hours and days following their release about one hour, then the containment vessel was isolated. from the bundle, and the effect of the presence of boric acid. Researchers focused on iodine radiochemistry in the sump water It was initially planned to introduce the hydrogen recombiner and the containment atmosphere, using several dedicated instru- coupons device in the containment atmosphere for 30 minutes ments. Painted surfaces installed in both the sump and the between the reactor shutdown and the end of steam injection, containment atmosphere (condensers) provided a source of but mechanical problems prevented this operation from being organic compounds capable of interacting with iodine. The performed(11). temperatures of the various vessel components were imposed The experimental phase then moved into a long-term phase to favor water evaporation from the sump and promote volatile lasting four days, with three successive stages: iodine transfer from the sump to the gas containment. an "aerosol" stage lasting around 37 hours, aimed at analyz- In addition to the objectives indicated previously, the tests also ing aerosol deposition mechanisms in the containment vessel. set out to characterize: During this stage, the vessel thermohydraulic conditions remained the size of aerosols received in the vessel and the aerosol unchanged; deposition processes (gravitational settling to vessel bottom, (9) Representative of situations where steam is almost completely transformed into (10) diffusiophoresis to the condensers, and deposition on elec- hydrogen by the Zircaloy oxidation reaction on the fuel cladding. tropolished vessel walls); (10) Aerosols are entrained by vapor condensation on the cooled parts of the condensers. any potential fission product poisoning effect on the cata- (11) For PHÉBUS-FPT3, the holder for the hydrogen recombiner coupons was lytic hydrogen recombiners used in nuclear reactor containments improved and introduced successfully [Biard et al., 2007].

IRSN - 2008 Scientific and Technical Report 93 2.3

a 23-minute washing phase, in which aerosols deposited by Hydrogen First Second oxidation gravitational settling on the hemispheric vessel bottom were concentration oxidation phase (vol. %) phase (fuel relocation) transferred to the sump. This was carried out by spraying the 100 2,500 sump water on the hemispheric vessel bottom through a loop 90 provided for this purpose; 80 2,000 70 a two-day chemistry stage, dedicated to analyzing iodine 60 1,500 chemistry in the sump and containment atmosphere, with 50 particular focus on iodine speciation. The water temperature 40 1,000 during this chemistry stage was brought from 90°C up to 120°C 30

Reactor shutdown Reactor to favor a 0.98 g/s evaporation/condensation cycle between 20 500 sump and condensers. 10 0 0 5,000 7,000 9,000 11,000 13,000 15,000 17,000 19,000 21,000 Time (s) t = 0 at 9:23 on October 12, 2000 Main results on bundle degradation Fuel bundle temperature at midplane (°C) ** Core power (a.u.) Hydrogen concentration in circuit (vol %) On-line measurements from the bundle, experimental circuit, and containment vessel during the reactor power increase detected the following events: Figure 2 General timeline of FPT2 degradation phase. Reactor the first cladding failure, which took place around the bundle power, fuel temperature at the 500 mm elevation, and hydrogen concentration in the circuit. midplane (roughly at the 406 mm elevation) at a temperature close to 800°C, as in previous tests; absorber rod failure occurred around the 500 mm elevation. The peak temperature measured on the guide tube at that time was about 1,350°C, close to the steel melting point (1,425°C, the temperature recorded a little later on the guide tube at 500 mm); the main oxidation phase occurred shortly after absorber rod failure in the bundle midplane, with a maximum bundle temperature of about 1,590°C at 500 mm. No significant material displacement was detected during the main oxidation phase, in contrast to FPT1, where oxidation was more acute; displacement of materials from the center of the bundle downwards was observed starting from the heating phase at a temperature of about 2,000°C and led to the formation of a molten pool in the lower portion of the bundle, as in previous tests; a chimney formed in the molten pool, allowing steam to escape. The chimney then plugged up, the molten pool spread Low density Very high density throughout the entire cross-section of the test bundle, and a significant temperature excursion was recorded, leading to reac- Figure 3 X-ray images of PHÉBUS-FP rod assembly before and tor shutdown. after FPT0, FPT1, and FPT2.

During FPT2, the hydrogen generation kinetics (Figure 2) were slower than in FPT0 and FPT1: the lower steam injection rate, approximately 0.5 g/s in FPT2 instead of about 2 g/s in FPT0 and FPT1, led to a slower progression of Zircaloy oxidation on the cladding. The oxidation front moved axially, first in the lower portion at a rate of 1.7 mm/s, then in the upper portion of the

94 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.3

bundle. Nonetheless, hydrogen generation lasted longer (an H2 a third significant period of release (of less volatile materials) concentration greater than 10% in volume is measured for 43 min was measured in the vessel during the second oxidation phase, utes) and led to a period of 18 minutes during which the hydro- corresponding to oxidation of the lower portion of the bundle, gen volume concentration in the circuit was between 75% and associated with the beginning of fuel relocation and 97%. liquefaction.

As in earlier tests, a second, less-significant oxidation stage After formation of the molten pool, no more significant amounts occurred in the final bundle degradation phase, when materials of fission products were received in the vessel, which does not flowed into the lower portion of the bundle. During this late imply, however, that release had stopped, since part of the oxidation phase, which lasted about 18 minutes (with a hydro- released materials could have been deposited in the circuit. gen volume concentration greater than 10%), a maximum hydrogen concentration of 17% was reached in the circuit for three The elements measured were classified according to the kinetics to four minutes. that led them to the vessel: release of noble gases started from the beginning of the first Based on the amount of hydrogen generated during degradation, oxidation phase. They accumulated in the vessel much more it was estimated that 81% of the Zircaloy in the cladding was quickly than the volatile fission products; oxidized. volatile and semi-volatile fission products (Cs, I, Rb, followed slightly later by Te and Mo) arrived in the vessel at an almost The final state of the bundle was observed on a post-test X-ray steady rate from the end of the first oxidation phase to the end image. Figure 3 shows the presence of a cavity at the center, of the second oxidation phase, which is consistent with a grad- with a molten pool in the lower portion consisting of a mixture ual degradation of the FPT2 bundle (Figure 4). This is also the of relocated, melted materials, with deformed rods in the upper case for indium, released by the absorber rod. Molybdenum was and lower portions. As expected, significant degradation of the only measurable in the vessel after the first oxidation phase, bundle was obtained during FPT2 (to approximately the same suggesting that release of this element was limited during the degree as that observed during FPT0, and to a greater extent hydrogen-rich phase and only became significant in the steam- than that observed during FPT1). rich phase that followed. This is consistent with the fact that oxidized molybdenum is more volatile than the metal; Total mass gain in the lower portion was evaluated at 3.9 kg the low-volatility fission products (Ba, La, Y, Sr), silver (mainly through non-destructive testing, 1.1 kg being attributed to from the absorber rod), elements from the structural components Zircaloy cladding oxidation, the rest corresponding to relocation and instrumentation (Zr, Re, W) and the fuel were mainly released of the materials. These observations were consistent with the during the end of the transient phase (second oxidation phase initial test objective in terms of degradation. and final heating period). The elements measured can be grouped together according to Main results from experimental circuit their fraction of global release from the fuel bundle: and containment vessel elements released in substantial amounts (around 80% of the bundle's initial inventory [hereafter noted i.i.]) such as noble Fission product release, transport, and deposition in gases (e.g. Xe); the primary circuit volatile fission products (I, Te, and Cs) with a released fraction During FPT2, as in the previous tests, release and transport of between 67% and 81% i.i. (Table 1); materials in the primary circuit correlated closely to events elements released in small or very small amounts, such as Ba occurring on the bundle. Three main phases of release were or Zr, which reached about 1.2% i.i. and 0.015% i.i., or even less identified: for other elements. the first significant release was observed during the first oxidation phase; The elements were also classified according to the amounts the second release phase was quasi-steady (observed for fission received in the vessel (referred to as the "vessel inventory"): products and the absorber rod elements) and continued practi- noble gases such as Xe and Kr, which did not react with surfaces cally up to the second oxidation phase; inside the circuit, reached the vessel without being retained in the circuit, so that the vessel inventory coincided with the amount

IRSN - 2008 Scientific and Technical Report 95 2.3

Ag-In-Cd First fuel Progression released from relocation of molten pool absorber rod Bundle ** cross-section ** First oxidation Progression blocked Normalized phase of fuel Cladding H2 concentration vessel relocation by-passes in circuit inventory steam (vol %) 1.2 120

1 100

0.8 99Mo 80

H2 generated 0.6 60

135Xe 0.4 131I 40 132Te 137Cs 0.2 20

0 0 7,500 9,500 11,500 13,500 15,500 17,500 19,500 Time (s) t = 0 at 9:23 on October 12, 2000

Figure 4 Volatile fission products received in the vessel and hydrogen concentration in the circuit.

released by fuel (about 79% i.i. and 60% i.i., respectively); The mass composition (Figure 5) of aerosols transported to the highly volatile fission products, particularly iodine (57% i.i.) vessel consisted predominantly of fission products (19% Cs, 20% and cesium (41% i.i.), behaved in a manner similar to that Mo), elements from the absorber rod (16% Ag, 15% Cd, 11% observed in FPT1; In), and an element contained in the cladding (7% Sn). Total for certain elements, a fraction ranging between 10% and mass was estimated at 55 g, taking into account material oxida- 40% of the initial inventory was received in the vessel (Rb (32% tion. This value was low compared to the 125 g(12) measured in i.i.), Mo (31% i.i.), Te (28% i.i.), Cd (23% i.i.), etc.). But signifi- FPT1 [Jacquemain et al., 2000; Dubourg et al., 2005]. cantly smaller amounts of Rb, Te, and Cd were transported to the vessel in comparison to FPT1, where they were more volatile, This significant difference can be explained partially by the lower with a vessel inventory of about 50% i.i.; steam injection rate used in FPT2 (four times less than FPT1). small amounts of the other elements were also received in This factor not only changed the degradation phenomena, hence the vessel, for example Sn (6.8% i.i.), In (5.7% i.i.), W (3.7% i.i.), the mass and composition of released elements, but also increased Ag (1.5% i.i.) and Tc (0.92% i.i.). Among these elements, the residence time in the bundle and the circuit, thereby favoring amounts of Sn, Ag, and Tc were considerably less than those deposition of certain elements in these areas. measured during FPT1, where the vessel inventory was 33% i.i., 6.7% i.i. and 21% i.i., respectively; Deposition negligible amounts (less than 0.5% i.i.) of a few elements were The processes whereby elements were deposited on the surface transported to the vessel, such as Sr, Ba, Ru, La, U, Zr, and Re. of the circuit lines include condensation of steam species in For most of these elements, the amounts in the vessel were thermal transition zones, impaction, thermophoresis, and aero- significantly less than those measured during FPT1. sol settling. Deposition was probably followed by chemical interaction with the surfaces or with other deposited elements. As in the previous tests, low amounts of barium were released, In FPT2, the low steam injection rate (0.5 g/s) resulted in sig- while separate effects tests seemed to indicate that the reduc- nificant deposition of volatile fission products (Cs, I, and Mo) in ing conditions and high temperatures would increase the released the upper portion of the bundle (Figure 6). This behavior was fraction of this element. FPT2 results therefore confirm the low volatility of barium, probably due to reactions between barium (12) Taking into account xenon and krypton, the mass received in the vessel was 150 g and the cladding materials [Dubourg and Taylor, 2001]. for FPT1 and 90 g for FPT2.

96 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.3

Release from Vessel bundle inventory (% i.i.) (% i.i.) Rb Te Others 2% 2% 5% Cs 67 41 Re I 72 57 3% Te 81 28 Sn Mo 52 31 7% 20% Mo

Table 1 Comparison of fractions of various fission products released In by the fuel bundle that reach the vessel. 11% Cs 19% Physicochemical properties of materials released Cd 15% in the circuit In the hot leg, significant fractions of cesium and iodine were Ag detected in the form of vapors(14) and aerosols. Measurements 16% taken on circuit samples after the FPT2 test showed that, first, vapor speciation changed throughout the different degradation phases, and second, cesium iodide was not the only iodine spe- Figure 5 Mass composition of aerosols received in the vessel during cies, contrary to common opinion. the whole FPT2 test. The condensed materials were either deposited on the walls of the steam generator tube, or transported to the cold leg as mixed not observed in FPT0 and FPT1, in which a higher steam injection aerosols. They were sized between 0.5 and 1 µm and were rate (2 g/s) led to significant deposits downstream (in the upper sometimes agglomerated. plenum above the bundle, and in the steam generator tube). Aside from tellurium, deposited in very large amounts in the The total transported aerosol mass was about 60 g in the hot upper plenum above the bundle and in the hot leg, and iodine, leg and 45 g in the cold leg (without taking into consideration deposited in the upper plenum and the vertical line in quantities the oxygen contribution in the oxides formed). The aerosol similar to those of FPT1 (although the proportions were inversed), concentration in the circuit reached its maximum during the deposits in the upper plenum, the vertical line, and the steam last oxidation phase and its mass composition changed as the generator tube were less significant in FPT2 than in FPT1 and various degradation phases took place: approximately 12% of FPT0. The total amounts deposited in the circuit are shown in the total mass went to the cold leg during the first oxidation Table 1. phase (predominantly the absorber rod materials, especially cadmium), 17% during the phase that led to fuel melting and After reactor shutdown, about 45% of the cesium deposited in relocation (mostly Cs and Mo volatile fission products, silver the upper plenum and in the vertical line was revolatized and coming primarily from the absorber rod and tin from the struc- deposited mainly on the surfaces of the steam generator tube. tural components), and 71% during the second oxidation phase These revaporization phenomena may be explained by a drop in (mostly silver, with significant amounts of volatile fission prod- the partial cesium pressure in the gas flow(13). Moreover, radio- ucts (Cs and Mo), absorbent materials such as Cd, and struc- active tellurium deposited in the hot leg was transformed into tural materials such as Sn). iodine through radioactive decay. The volatile iodine was also transported to the steam generator tube. These observations indicate that in the event of a hot leg break, tellurium and (13) Corresponds to steam injection, maintained for one hour after reactor shutdown to cool the fuel bundle. cesium deposits in the hot leg could lead to additional release (14) In this context, vapors refer to the compounds that condense between 150°C of volatile forms of iodine and cesium, respectively. and 700°C, as opposed to gases, which do not condense.

IRSN - 2008 Scientific and Technical Report 97 2.3 FPT 2

Linear activity (i.i. %/mm) Upper rods Cavity Molten pool Grid + upper rods 2.10-1 2.10-1 2.10-1 1.10-1 FPT2 1.10-1 1.10-1 8.10-2 FPT1 6.10-2 4.10-2 2.10-2 0 1,100 1,000 900 800 700 600 500 400 300 200 100 0 -100 Elevation (mm from bottom of fissile column) FPT1

Low density Very high density

Figure 6 Comparison of distribution profiles of 137Cs in the fuel bundle in FPT1 and FPT2.

Aerosol behavior in the vessel condensation rate that was four times lower in FPT2 (0.5 g/s Inside the vessel, aerosols were subject to three deposition instead of 2 g/s in FPT0 and FPT1). Measurements showed non- phenomena (March et al., 2007; March 2008): negligible deposits on the vessel walls (about 11% of the vessel diffusiophoresis, corresponding to aerosol entrainment through inventory), higher than FPT1 (2% at most). These differences in steam condensation on the condensers(15); deposition can be partially explained by the aerosol size, which gravitational settling at the bottom of the vessel (in the sump was lower in FPT2 than in FPT1. and on the hemispherical vessel bottom); deposition on all vessel walls. At the end of the aerosol phase, the vessel inventory fractions deposited on the vessel walls and bottom were practically the Most of the isotopes detected using gamma spectrometry showed same for all the elements, whereas the distribution between the the same overall behavior. Gravitational settling was the pre- sump and the condensers depended on element solubility. The dominant mechanism in the vessel: on an average, 74% of the soluble elements (mainly cesium and rubidium) were almost all vessel inventory was deposited at the bottom of the vessel. found in the sump, while insoluble elements were found in both Diffusiophoresis entrained 12% of the vessel inventory to the the sump and on the condensers. condensers. The aerosol deposition rate on the condensers was, like the condensation rate, four times less than that observed (15) Condensates are recovered at the bottom in a capsule that empties into the in FPT0 and FPT1, which is consistent with a steam injection/ sump as soon as the level reaches a given threshold.

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During the degradation phase and the beginning of the aerosol from that observed during FPT0 and FPT1, where vapor conden- phase, aerosol solubility led to three main categories: sation and aerosol thermophoresis were both observed, leading elements found mainly in the aqueous phase and considered to larger deposits, i.e. 23.5% i.i. for FPT0 and 19.2% i.i. for FPT1. as soluble (Cs, Rb, and I); In FPT2, the lower steam injection rate may have led to deposi- elements partially soluble in water (Ba, Mo, Cd, Re, and Tc); tion that was not measured, through condensation upstream elements that remain practically insoluble (Ce, Te, Zr, Ru, Sn, from the steam generator tube. In, Ag, W, and U). This behavior was equivalent to that observed during FPT1 (with an acid sump), except for iodine, which, in In the cold leg, most of the iodine was transported as aerosols FPT1, behaved like a soluble element during degradation, then and about 57% i.i. reached the vessel, which is slightly less than like an insoluble element during the aerosol phase. in FPT0/FPT1 (about 63% i.i.) [Clément et al., 2006; Girault et al., 2006; Jacquemain et al., 1999]. In the containment atmosphere, aerosol size showed a ten- dency to increase during the degradation phase, with a mean Iodine behavior in the containment vessel mass aerodynamic diameter that increased from 1.4 µm in the During the degradation phase, iodine that reached the vessel first oxidation phase to 3.68 µm at the beginning of the aerosol was deposited mainly through gravitational settling on the phase. bottom of the vessel and in the sump (74% of vessel inventory), while the rest was either entrained through diffusiophoresis to Iodine release and behavior in the circuit the condensers (11% of vessel inventory), or deposited on the Iodine released from fuel during the degradation phase was vessel walls (11% of vessel inventory). This behavior is typical estimated at about 73% i.i., which was lower than FPT1 and of aerosols. The solubility of iodine in the sump (approx. pH 9) FPT0 (about 87% i.i.). This difference can be explained by the changed during the course of the test. Before the hemispheric significant deposits observed in the upper portion of the bundle. bottom of the sump was washed, the iodine collected in the This retention phenomenon was related to the lower tempera- sump appeared primarily as soluble species or colloidal suspen- tures measured in this zone, due to the lower steam injection sions. The iodine brought in during the washing phase generate rate (four times lower than in FPT1 and FPT0). Moreover, in the an insoluble iodine fraction, reducing the soluble iodine fraction upper plenum and the vertical line, iodine probably reacted with to 66%. This phenomenon can be explained by two mecha- fission products (Cs) and absorber rod materials (Ag, In, and Cd) nisms: to form metal iodine vapors, leading to significant deposition: during the washing phase, soluble iodine and insoluble ele- 4.1% i.i. in the upper plenum and 0.4% in the vertical line, ments (silver particles) were added simultaneously to the sump; respectively (in comparison, in FPT1 deposition was 1.5% and iodine was then able to react with these elements to form 3.8%, respectively), where large temperature drops were mea- insoluble iodine species; sured on the walls. In the hot leg, as in previous tests, iodine the iodine species deposited on the hemispherical vessel was found mainly in the vapor form, at 700°C, gaseous iodine bottom may then have reacted chemically with other elements representing only 0.1% i.i. and iodine in aerosol form represent- present in the containment atmosphere to form insoluble iodine ing between 10% i.i. and 44% i.i., depending on the degradation species, which were then entrained to the sump during the phase, which was higher than FPT1 results (less than 15% i.i.). washing phase.

During the degradation phase, various chemical forms of the In the chemistry phase of FPT2, the soluble form of the iodine iodine transported as vapor through the hot leg of the primary fraction was much higher than in the previous tests (iodine was circuit were revealed, during analysis of the condensates depos- almost always insoluble in FPT1). This difference can be explained ited on the walls of the sampling lines and the thermal gradient by the lower Ag/I molar ratio (only 10, whereas it was 50 in FPT1 tube in a zone where the temperature dropped from 700°C to and 2,000 in FPT0), slowing down the kinetics of the reaction 150°C. At least two species were distinguished: cesium iodide, between Ag and I, and by the alkaline state, which did not favor detected only after the main oxidation phase, and a second, the formation of the insoluble specie AgI [Funke et al., 1996]. more volatile specie that remains unidentified at this time. During In FPT2, the gaseous fraction of iodine measured in the vessel FPT2, about 6.2% i.i. was deposited on the steam generator tube, reached its peak (about 0.3% (i.i.) of the iodine mass initially with an exponential deposition profile characteristic of aerosol present in the bundle), during the second oxidation phase, deposition through thermophoresis. This behavior was different whereas this peak was reached in the first oxidation phase in

IRSN - 2008 Scientific and Technical Report 99 2.3

FPT0 and FPT1. Changes in the gaseous iodine fraction in the Gaseous iodine fraction in vessel (% of initial inventory) containment atmosphere over the long term are shown in 0.35 Figure 7. Aerosol phase Chemistry phase 0.3 Aerosol phase Chemistry phase As in the previous tests, a sharp drop in the gaseous iodine 0.25 fraction in the containment atmosphere was observed after the second oxidation phase (from 0.3% i.i. to 0.1% i.i.), suggesting 0.2

that iodine was efficiently trapped on the cooled painted sur- 0.15 faces. During the aerosol phase, the gaseous iodine fraction 0.1 reached a plateau of about 0.1 to 0.15% i.i., probably correspond- 0.094± 0.019 % ing to a physicochemical iodine equilibrium reached inside the 0.05 0.055± 0.012 % vessel. After the washing phase, the gaseous iodine fraction 0 0.010± 0.024 % decreased by about one order of magnitude, reaching 0.01% i.i. 30,000 80,000 130,000 180,000 230,000 280,000 330,000 380,000 at the end of the chemistry phase. This decrease was not observed Time (s) in FPT1, where the gaseous iodine fraction was doubled after FPT1 FPT2 the washing phase, reaching equilibrium at about 0.2% i.i. The difference is attributed to alkalinity in the sump (approx. pH 9) and the evaporating conditions chosen for FPT2, whereas FPT1 Figure 7 Evolution of the gaseous iodine fraction in the containment took place with a non-evaporating, acid sump (approx. pH 5). atmosphere over the long term for FPT1 and FPT2. During the chemistry phase, a decrease in the iodine deposits on vessel surfaces was observed, where about 0.9% i.i. and 0.2% i.i. of the iodine was desorbed from the vessel walls and the condenser surfaces, respectively. In the latter case, the desorption Outlook mechanism was not related to gaseous desorption, but was most likely the result of condensation washing the condensing sur- On an overall basis, the FPT2 objectives were reached satis- faces, since this decrease in deposition was also observed for factorily [Gregoire et al., 2008] and the numerous Mo and Te, present only in aerosol form. In FPT0, FPT1, and FPT2, experimental results obtained either confirmed experimental the gaseous iodine fraction during the chemistry phase was low observations made during previous tests (for example, the (less than 0.1% i.i.) because it was retained in the sump. iodine trapping effect produced by silver in the sump due to the formation of insoluble AgI), or contributed new findings In FPT0 and FPT1, the trapping mechanism was explained by the (such as the discovery that CsI is not the primary iodine spe- presence of silver (mainly from the absorber rod), whose effect cie transported through the circuit). FPT3 [Biard et al., 2007], was enhanced by the acid pH. In FPT2, consumption of gaseous conducted successfully from November 18 to 22, 2004, iodine was enhanced, first, by the increase in mass transfer brought the PHÉBUS-FP experimental program to a close. between the sump and the containment atmosphere (favored The special feature of the FPT3 test was that boron carbide by an evaporation/condensation cycle of 0.98 g/s), and second, was used as the boron absorber, instead of the silver-indi- by the hydrolysis reactions in the sump, under alkaline (approx. um-cadmium material. Experimental studies on severe pH 9) and high-temperature (120°C) conditions. accidents in nuclear water reactors will continue at IRSN with the international Source Term program [Clément and Throughout FPT2, the gaseous iodine fraction was predominantly Zeyen, 2005b], consisting of separate-effect testing. inorganic (80 to 90% of total gaseous iodine), in contrast to FPT0 and FPT1, where organic iodine was the major To assess environmental release in the event of a severe acci- component. dent, these results must be transposed to a power reactor scenario investigated using numerical simulation tools. This work is currently under way.

100 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.3

References B. Biard, Y. Garnier, J. Guillot, C. Manenc, P. March, F. Payot (2007). FPT3 preliminary report, Document PHÉBUS-PF IP/06/569. J. Birchley, T. Haste, H. Bruchertseifer, R. Cripps, S. Güntay, B. Jäckel (2005). Phébus-FP: Results and significance for plant safety in Switzerland, Nuclear Engineering and Design, vol. 235, p. 1 607-1 633. P. Bottomley et al. (2007). Post irradiation examination of the lower part of the PHÉBUS-FPT2 degraded bundle, International Congress on Advanced Power Plants (2007), Nice, France, May 13-18. B. Clément (2003a). Summary of the PHÉBUS-FP Interpretation status, Proc. 5th Technical seminar on the PHÉBUS-FP programme, Aix-en-Provence, France, June 24-26. B. Clément, N. Hanniet-Girault, G. Repetto, D. Jacquemain, A.-V. Jones, M.-P. Kissane, P. Von der Hardt (2003b). LWR severe accident simulation: synthesis of the results and interpretation of the first PHÉBUS-FP experiment FPT0, Nuclear Engineering and Design, vol. 226, p. 5-82. B. Clément, R. Zeyen (2005b). The PHÉBUS Fission Product and Source Term international programs, International Conference Nuclear Energy for New Europe 2005, Bled, Slovenia, September 5-8. B. Clément, N. Girault, G. Repetto, B. Simondi-Teisseire (2006). Les enseignements tirés du programme PHÉBUS-PF, RST IRSN 2006, 84-96. R. Dubourg, R.P. Taylor (2001). A qualitative comparison of barium behaviour in the PHÉBUS-FPT0 test and analytical tests, Journal of Nuclear Materials, vol. 294, p. 32-38. R. Dubourg, H. Faure-Geors, G. Nicaise, M. Barrachin (2005). Fission product release in the first two PHÉBUS tests FPT0 and FPT1, Nuclear Engineering and Design, vol. 235, p. 2 183-2 208. J.-M. Evrard, C. Marchand, E. Raimond, M. Durin (2003). Use of PHÉBUS-FP Experimental Results for SOURCE TERM Assessment and Level 2 PSA, Proc. 5th Technical seminar on the PHÉBUS-FP programme, Aix-en-Provence, France, June 24-26. F. Funke, G.-U. Greger, A. Bleier, S. Hellmann, W. Morell (1996). The reaction between iodine and silver under severe PWR accident conditions, Chemistry of Iodine in Reactor Safety, Workshop proceedings Würenlingen, Switzerland 10-12 June, NEA/CSNI/R(96)6. N. Girault, S. Dickinson, F. Funke, A. Auvinen, L. Herranz, E. Krausmann (2006). Iodine behaviour under LWR accidental conditions: lessons learnt from analyses of the first two PHÉBUS-FP tests, Nuclear Engineering and Design, vol. 236, p. 1 293-1 308. A.-C. Gregoire, P. March, F. Payot, B. Biard. A. de Bremaecker, S. Schlutig (2008). FPT2 final report, Document PHÉBUS IP/08/579. D. Jacquemain, N. Hanniet, C. Poletiko, S. Dickinson, C. Wren, D. Powers, E. Krausmann, F. Funke, R. Cripps, B. Herrero (1999). An Overview of the Iodine Behaviour in the Two First Phébus Tests FPT0 and FPT1, OECD Workshop on Iodine Aspects of Severe Accident Management, Vantaa, Finland, May 18-20. D. Jacquemain, S. Bourdon, A. de Bremaecker, M. Barrachin (2000). FPT1 final report, Document PHÉBUS IP/00/479. P. March, F. Payot, B. Simondi-Teisseire (2007). Fission product behaviour in the containment during the FPT2 test, MELCOR Cooperative Assessment meeting, Albuquerque (NM), USA, September 18-20. P. March (2008). Fission product behaviour in the PHÉBUS containment, American Nuclear Society: 2008 Annual Meeting, Anaheim, USA, June 8-12. M. Schwarz, G. Hache, P. Von der Hardt (1999). PHÉBUS-FP: a severe accident research programme for current and advanced light water reactors, Nuclear Engineering and Design, vol. 187, p. 47-6Z. M. Schwarz, B. Clément, A.-V. Jones (2001). Applicability of PHÉBUS-FP results to severe accident safety evaluations and management measures, Nuclear Engineering and Design, vol. 209, p. 173-181. J.-P. Van Dorsselaere, H.-J. Allelein (2004). ASTEC and SARNET – Integrating Severe Accident Research In Europe, Proc. EUROSAFE Forum, Berlin, Germany, 2004.

IRSN - 2008 Scientific and Technical Report 101 2.4 MEPHISTA: a thermodynamic database on nuclear fuel developed for and applied to nuclear safety

Marc BARRACHIN Corium and Radioelements Transfer Research Laboratory

Nuclear fuels are complex materials, subjected to specific environmental constraints (irradiation in normal operation, oxidizing environment and high temperature in accident conditions, etc.). To understand their behavior in normal operation and in accident conditions, it is necessary to call on several disciplines that include thermodynamics, solid physics, mechanics, and others.

Among these fields, thermodynamics plays a very particular role. It provides knowledge on the arrangement of the atoms of a material (in other words, its phases at the thermodynamic equilibrium), the manner in which this arrangement is modified depending on parameters, such as composition and temperature, and subsequently, the conditions in which a transformation may occur in a given direction. Of course, it does not say anything about the transformation mechanisms themselves, or their duration, and thus says nothing about kinetics of this transformation. But it is important to make a distinction between those transformations that are possible, and those that are not, which is what thermodynamics is capable of achieving with certainty. From this point of view, thermodynamics is the focal point of all the disciplines called on to describe fuel behavior.

Investigating the transformations of a nuclear fuel in terms of Under normal irradiation conditions in a PWR, fission products thermodynamics only makes sense if the object of the study are found inside the fuel matrix in different phases and differ- covers both the elements contained in the fuel (or fuel matrix, ent physical states (condensed or gaseous) [Kleykamp, 1985]: for example uranium and oxygen for the uranium dioxide fuel in the form of oxides dissolved in the matrix for nearly half used in pressurized water reactors, PWRs) and the fission prod- of them: Sr, Y, Zr, La, Ce, Nd, etc.; ucts generated by irradiation. It is essential to understand the as oxide phases: Ba and Nb; thermodynamics of the entire system for various reasons as metallic phases: Mo, Ru, Tc, Pd, Rh; related not only to fuel safety (as in the case of cladding dam- in gaseous form: Br, Rb, Te, I, Cs; age caused by a chemical interaction between cladding and the in the form of atoms dissolved in the matrix or fission gas irradiated fuel in a PWR component), but also fuel performance bubbles: Xe and Kr. (such as the impact of the different phases of irradiation on fuel properties).

102 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.4

Difficulties arise due to the fact that the chemical nature of certain fission products, especially those belonging to the first Theoretical Estimates Experiments calculations three categories, is not permanent and can change depending Differential thermal Quantum analysis, calorimetry, on the operating temperature, the oxygen potential in the fuel, mechanics steam pressure measurement, X-ray and the burnup. Statistical Models diffraction, etc. thermodynamics with adjustable parameters In a severe accident (involving partial melting of fuel matrix components), the role of fission products must be examined Thermodynamic Parameter modeling adjustment from a different angle. The most volatile (Xe, I, Cs, Te) are released calculation s from the reactor core when it is degraded, and form part of the ab initio ab radioactive substances that may potentially be released from Thermodynamic function the facility. Other less volatile fission products (Zr, La), some G, F… of which release heat, may remain trapped inside the fuel and contribute to heating the core, thereby aggravating core deg- Thermodynamic Minimization and/or database equilibrium radation. In this situation, a thermodynamic study of the fuel must include not only the chemical reactions between the fuel Equilibrium calculations Phase Graphic and fission products, but also any possible interaction between diagram representation the degraded fuel and its cladding (and certain structural mate- Applications rials, as necessary), for various gaseous atmospheres representative of accident scenarios, and for temperatures that may Figure 1 Schematic description of the CALPHAD method, used to range up to the fuel melting point. develop the MEPHISTA thermodynamic database.

It quickly becomes apparent that this presents an extremely complex chemical system, especially in terms of scope (i.e. the number of chemical elements to be considered), and the pos- considered and the wide temperature range to be covered. It is sible associations that may be formed between the various easy to understand that an experimental approach alone cannot chemical elements (through the formation of compounds, for meet this challenge, even if it remains essential, as will be example). It is therefore necessary to work with a thermody- seen. namic model that covers a sufficient number of chemical elements to accurately describe the thermodynamic behavior of Another approach consists of building a phase diagram through each one. calculation. The most fundamental approach consists of writing the thermodynamic partition function (Z) representative of the microscopic properties of the material, and studying the sin- Approach to building thermodynamic gularities of Z to determine the phase boundaries. Z provides databases access to all the thermodynamic quantities (for example, the Gibbs energy is written G = -kTLnZ at the thermodynamic Conventionally, thermodynamic knowledge of a material is limit). ascertained by establishing a phase diagram, i.e. a graphic representation of the arrangement of the atoms of the mate- Calculation (even approximate) of a partition function is cer- rial, which generally depends on its composition and tempera- tainly the most difficult problem to solve in thermodynamics. ture. The phase diagram is determined experimentally by It requires having the appropriate model to express the internal measuring different properties (phase-change temperature, energy of the system (generally based on its electronic structure) phase composition after quenching, etc.). Compendiums list and an approximation to calculate entropy (mean field, Monte these diagrams, established experimentally for simple materials, Carlo simulation, etc.) [Finel, 1987; Barrachin, 1993]. This meth- binary systems (i.e. compounds formed by two chemical ele- odology, already quite complex for a binary system, quickly ments, as in [Hansen and Anderko, 1991]), and certain ternary becomes unworkable when considering materials consisting of systems (three elements). For fuel, the task is much more more than three chemical elements. complex, given the large number of chemical elements to be

IRSN - 2008 Scientific and Technical Report 103 2.4

For complex materials, the selected approach consists of deter- especially measured thermodynamic quantities, that determines mining the macroscopic thermodynamic behavior of each phase the significance level assigned to the thermodynamic model of that may appear in the phase diagram. For a constant pressure the material in question. For the MEPHISTA database, this phase diagram, this behavior may be described using the Gibbs significance level translates into a quality level associated with energy, G. The mathematical form of G for a given phase is the model of each binary and ternary system. generally a function of composition and temperature. To write the function explicitly, it is usually necessary to call on models designed to describe the microscopic reality of the material as Models to describe Gibbs energy in closely as possible, while being careful to express G in a rela- different phases tively simple manner. To set the CALPHAD method to work, it is first necessary to The coefficients used in these models are determined from describe how Gibbs energy, G, in the different phases depends experimental data, or based on assumptions when data is not on temperature and composition. This dependence is a function available. Given the Gibbs energies for the different phases, it of the type of phase: a distinction is made between pure ele-

is possible to determine the "Gibbs energy function" for all the ments (U, O, etc.), stoichiometric compounds (U3O8, ZrO2, etc.)

composition and temperature domains, and by minimizing this and solutions (solids, for example UO2±x , or liquids). functional, to obtain the phase stability domains, i.e. the phases that may occur for a given composition and temperature. For elements and stoichiometric compounds, the fundamental thermodynamic quantities are the standard enthalpy of forma- This method, called the CALPHAD method (Calculation of tion ( ), entropy at ambient temperature ( ) and phase diagram), described for the first time by [Kaufman and the heat capacity as a function of temperature at constant Bernstein, 1970] (Figure 1) at the end of the 1960's, can be pressure . Based on this data, G can be calculated using used to merge the phase equilibrium thermodynamics of J. W. the following classical equations: (a) Gibbs and numerical calculation. This mixed approach, based on both calculation and experimentation and widely used to where: describe complex materials, was employed to describe fuel (b) thermodynamics in the MEPHISTA database.

Contrary to the "fundamental" approach described previously, the CALPHAD method is only partially predictive, since construc- The analytical form used to express heat capacity at constant tion of the Gibbs energies for the various phases is based on pressure leads to a general equation for G(T) expressed adjusting models to reproduce experimental results. It nonethe- as: less offers the advantage of being able to predict the thermodynamics of a complex material based on modeling of the For solutions that may be stable over a wide concentration binary and ternary systems alone, making it a very powerful domain, G is developed as a function of temperature T and

method. It can also be used to integrate experimental data composition c = (ci)i = 1,N (N being the number of constituents deduced from phase diagrams (mainly stability domain bound- in the solution). aries and transition temperatures) and data obtained by measur- The double dependence of G is generally taken into account by ing thermodynamic quantities (enthalpy of formation, activity, calculating the sum of two terms: chemical potential, etc.) in a single function (G), thereby ensur- G(c,T) = Gref (c,T) + Gmix (c,T) ing consistency between these various sources of information. The first term, Gref(c,T), is the weighted sum of the Gibbs ener- It is nonetheless important to remember that these thermody- gies of the solution constituents. namic quantities unequivocally determine the energy level in the Gibbs energies of the various phases, which cannot be achieved using only the topology data from the phase diagram. This is a very important point to be considered when attempt- ing to model the thermodynamics of a complex material. From this point of view, it is the abundance of experimental data,

104 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.4

U κ 4,400

4,000 O 3,600 (1963Blu) (1965Mar) 3,200 G L1 L2 (1966Gui) 2,800 (1967Ban) (1968Kot) 2,400 (1969Ack) 2,000 (1970Lat) (1970Tet) 1,600 (1974Sai) 1,200 (1975Mat) 1980Gar) 800 ( (1998Gue) 400 (2005Man) 2005Man) 0 ( 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 O U Molar fraction a

Figure 2 Crystalline structure of uranium dioxide: the uranium atom Figure 3 Calculated O-U phase diagram (MEPHISTA shown as solid sublattice is cubic with centered faces, while the oxygen line) compared with experimental points (see Figure 6 also atoms are arranged in a simple cubic sublattice with a for a close-up view of UO2 composition at high temperature). mesh of a/2.

The second term, Gmix(c,T), corresponds to the solution's Gibbs Nonetheless, each of these contributions can take a noticeably energy of mixing. The difficulty lies in estimating this mixing different form, depending on the model. term, for which there are different models, each new model To describe the thermodynamics of oxide fuels, it is particu- being developed to improve the reliability of the extrapolation larly important to carefully choose the models relative to the of G to temperature and composition domains for which no solution phases present in the O-U binary system (the principal experimental data is available. system to be modeled). Uranium dioxide, UO2, crystallizes in

the fluorite-type cubic structure (CaF2) (Figure 2).

In general, the Gibbs energy of mixing is expressed as the sum Atom ordering in the non-stoichiometric solid solution UO2±x of two terms, the first corresponding to the Gibbs energy of can be described using a three-sublattice model with defects ideal mixing (mixing "without excess interaction"(1) between (see [Sundman and Agren, 1981]), the defects making it pos- the different constituents in the solution, i.e. the perfect gas sible to cover the entire solution composition domain using the hypothesis), the second expressing the difference between the same model. The first sublattice contains the uranium atoms, Gibbs energy of mixing and this hypothetical ideal (referred to the second the oxygen atoms and vacancy-type defects used as the "excess Gibbs energy of mixing"): to describe the sub-stoichiometric composition domain (UO2-

x), and the last sublattice contains interstitial oxygen atoms for

UO2+x. This representation does not correspond strictly to where: microscopic reality as described by [Willis, 1964], but it makes it possible to represent the experimental data obtained to establish the phase diagram (Figure 3) and the measured ther- The excess Gibbs energy of mixing is usually approached using modynamic quantities (Figure 4), which are quite abundant for polynomials given by [Redlich and Kister, 1948]: the U-O system.

For the liquid solution, Gibbs energy is calculated using the The above expressions are very general, making it possible to associated model, developed by [Prigogine, 1950], which assumes separate the different contributions (reference, ideal, and excess). non-ideal mixing the between uranium and oxygen atoms and the uranium dioxide molecules. This description covers the entire composition domain, from (1) "Without excess interaction" means that the interaction force between atoms of the liquid solution of the metal (pure uranium) to the over a different type is about half the sum of the interaction forces between atoms of the same type. stoichiometric oxide (UO2+x).

IRSN - 2008 Scientific and Technical Report 105 2.4

0 Temperature (K) Reflectivity, arbitrary units 1,573 K -50,000 3,400 3

3,300 -100,000 2.5 3,200

2 -150,000 3,100 T liquidus

3,000 1.5 -200,000 T solidus

1 -250,000 2,800 0.300 0.305 0.310 0.315 0.320 0.325 0.330 0.5 Markin et al. 1962 2,700 Aukrust et al. 1962 Hagemark and Broli 1966 2,600 0 Lindemer and Besmann 1985 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 1.1 Labroche 2003 Time (s) MEPHISTA

Figure 4 Oxygen potential 1/2 DG02 in kJ/mol, calculated at 1573 K for Figure 5 Flash laser method used to determine liquidus and solidus the solid solution Ux(U)O2/3 as a function of x(U), compared temperatures of the UO2-ZrO2 system for a composition of with experimental points. 80 mol % UO2, 20 mol % ZrO2, COLOSS program [Ronchi and Sheindlin, 2002].

MEPHISTA thermodynamic database To give an idea of the enormous task accomplished, building the MEPHISTA database required modeling the Gibbs energy of The MEPHISTA database (Multiphase Equilibria in Fuels via 49 solution phases, 219 pure elements or condensed stoichio- Standard Thermodynamic Analysis) [Cheynet and Fischer, 2006] metric compounds, and 151 gaseous compounds. has been under development since 2000, and aims to describe fuel thermodynamics. This work is being conducted through a The model can process irradiated fuel thermodynamics for

collaborative project between Thermodata, INPG (French products as varied as uranium dioxide (UO2) and the MOX fuels

National Physics Institute of Grenoble), CNRS (French National used in PWRs, or TRISO particles (UO2 or UO2-UC2 fuel) which Scientific Research Center), and IRSN. The database contains are expected to fuel 4th generation high-temperature reactors. the Gibbs energy coefficients of phases that may be formed in The MEPHISTA database also serves to develop applications for the O-U-Pu-Zr-Fe-Si-C-Ba-Cs-La-Mo-Sr-Ru chemical system. carbide fuels that may be used in gas-cooled fast reactors (GFR) Oxygen (O), uranium (U) and plutonium (Pu) are the elements and for oxide fuels used in sodium-cooled fast reactors (SFR). used to produce oxide fuels; zirconium (Zr, used for fuel cladding, but also considered as a fission product), barium (Ba), cesium In the coming years, the MEPHISTA database will be enriched (Cs), lanthane (La), molybdenum (Mo), strontium (Sr), and with chemical elements to complete the list of fission products, ruthenium (Ru) are the fission products that are important in resulting in a more detailed description of oxide fuel thermo- describing the thermodynamics of irradiated fuel. dynamics, especially with regards to fission products involved In the MEPHISTA database, iron (Fe) is included to describe in what is referred to as the metal phase of Pd-Tc-Ru-Mo-Rh. interactions between the fuel rod and certain structural mate- The database may also be extended to include doping elements rials (reactor internals, control rod cladding, etc.) in accident currently planned for use in oxide fuels, such as gadolinium and conditions. The chemical elements silicium (Si) and carbon (C) chromium. were added to study the thermodynamics of TRISO fuel on high-temperature reactors [Phélip et al., 2006]. Lastly, argon and hydrogen are included only in modeling the Gibbs energy of gaseous compounds.

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Programs to support MEPHISTA database Calculated Measured development Cs2MoO4 -1664,9 -1514,5 ± 1.0

Cs UO -1879,9 -1928,2 ± 1.0 As mentioned previously, the quality of the CALPHAD approach 2 4 depends primarily on the experimental measurement data Cs2ZrO3 -1587,9 -1584,8 ± 1.0 available to determine the Gibbs energy coefficients. Most Table 1 Comparison of standard enthalpies of formation obtained metallic binary systems have been studied experimentally, but by calculating the electron structure [F. Gupta, 2008] and measured enthalpy [Cordfunke and Konings, 1990] (in data on oxide binary systems, particularly those containing kJ/mol). uranium and plutonium oxides, are not as extensive. The amount of experimental data on ternary and quaternary systems is even less. laborative network (currently being organized), with the data- For more than ten years, a vast experimental effort has been base developed at the Royal Military College of Canada. undertaken through various national and international programs Another source of information consists of the results obtained to fill in these gaps. IRSN is extensively involved in these pro- from ab initio calculations, which involve resolving quantum grams, which include CIT (a European Commission (EC) project), mechanics equations. Without entering into a detailed theo- COLOSS (EC) (example of determination of liquidus and solidus retical discussion, it should be noted that this approach can be temperatures in the UO2-ZrO2 system, Figure 5), RASPLAV (an applied to obtain values on the enthalpy of formation ( ) OECD/CSNI(2) project), MASCA (OECD/CSNI), MASCA2 (OECD/ for stoichiometric compounds to a satisfactory degree of accu- CSNI), ENTHALPY (EC) and CORPHAD (an ISTC(3) project). racy. These values can contribute to the determination of Particular focus has been given to the U-O-Zr-Fe quaternary coefficients used to express Gibbs energy (equations (a) and system [Chevalier et al., 2004; Bechta et al., 2006; Bechta et (b)). For example, Table 1 shows a comparison of the standard al., 2007], a key system for describing fuel rod degradation in enthalpies of formation obtained by calculating the electron accident conditions. structure and measured enthalpy values for certain stoichio- This effort is being pursued today in the ISTC PRECOS program, metric compounds of cesium. in which IRSN plays an important role. The MEPHISTA database is gradually incorporating new experimental data obtained through the above programs, as well as Applications in nuclear safety information published in open technical literature, by perform- ing regular updates on coefficients used to express Gibbs The MEPHISTA database and the NUCLEA database, specifi- energy, thus contributing to a constant improvement of the cally dedicated to corium thermodynamics, are used by a large significance level. In this aspect, it is a unique tool for building number of institutes, industrial partners, and universities, includ- knowledge on nuclear fuel thermodynamics. ing EDF, CEA, Areva, VTT (Finland), AEA-T (Great Britain), In spite of these continuous efforts, there are composition and Alexandrov RIT (Russia), KAERI (South Korea), AECL (Canada), temperature domains for which there is no experimental data. Risø National Laboratory (Denmark), Boise State University Phase-equilibrium thermodynamics being a relatively old dis- (USA), and others. cipline (dating back to the 19th century), whatever has not been At IRSN they are used either coupled to applications developed accomplished so far is not easy to achieve. Calling on modeling at the Institute to simulate core degradation in accident condi- hypotheses to build Gibbs energy data for certain phases can tions (for example, the ICARE/CATHARE code [Salay and Fichot, open a new road to this information. In this context, it is impor- 2005]), or to support interpretation of various test results (such tant to confront these hypotheses with other alternatives. This as those from tests conducted in the OECD MASCA programs confrontation can take place by comparing thermodynamic [Barrachin and Defoort, 2004]). databases. In 2003, as part of the ENTHALPY program [Debremaecker et al., 2003], the MEPHISTA database was com- (2) CSNI: Committee on the Safety of Nuclear Installations. pared with the THMO database established by AEA-Technology (3) ISTC: International Science and Technology Center. (Great Britain), with the aim of reaching a consensus on mod- (4) The SAMANTHA collaborative network (Simulation by Advanced Mechanistic eling hypotheses [Cheynet et al., 2004]. There are plans to and Thermodynamic Approaches to nuclear fuels), coordinated by IRSN, aims mainly to develop detailed models describing fission product behavior in nucle- pursue this exercise in the context of the SAMANTHA(4) col- ar fuel.

IRSN - 2008 Scientific and Technical Report 107 2.4

The following discussion presents two recent applications of the MEPHISTA database, the first involving interpretation of O, U

test results from the PHÉBUS-FP program, focusing on estima- T/K - Temperature tion of the temperature at which loss of integrity occurs on fuel -3,200 rods under accident conditions in a pressurized water reactor -3,000

[Barrachin et al., 2008], the second having a more predictive -2,800 objective, since it is dedicated to fission product behavior in FCC_C1 -2,600 TRISO fuel, which may be used in 4th generation very-high -2,400 temperature reactors (VHTR) [Barrachin et al., 2008]. -2,200

Fuel failure temperature under severe accident -2,000 conditions 0.250 0.270 0.290 0.310 0.320 0.350 0.370 0.390 0.410 0.430 0.450 OU In a hypothetical core meltdown accident in a pressurized water X – Molar fraction reactor caused by loss of coolant and unavailable safeguard (1970Lat) 2005Man) systems, interaction between the Zircaloy-4 cladding and the ( (2005Man) fuel (UO2) is the main mechanism leading to fuel rod degrada- MEPHISTA – tion, thus core degradation. Fuel can be subjected to various gaseous environments (steam and hydrogen), which determine Figure 6 O-U phase diagram (high-temperature portion, close-up of Figure 3), compared to experimental points obtained by the type of interaction between the cladding and the fuel, and [Manara et al., 2005]. the temperature at which fuel integrity will be lost. Fuel exposure conditions change throughout the course of an accident scenario: fuel may be exposed to temperatures ranging from [Clément et al., 2003; Jacquemain et al., 2000]. These tests, 900 K to 2,800 K, and to varying atmospheres, from a high- conducted in an oxidizing atmosphere (at a pressure of 0.2 MPa)

oxidation environment (steam in excess), to a high-reduction with UO2 fuel, unirradiated for FPT0, irradiated to 2.4 at % for environment (rich in hydrogen). FPT1, revealed that fuel collapse occurred near 2,500 K, i.e. 300 K lower than what was generally expected. In a reducing environment, the chemical affinity of Zircaloy for oxygen will extract oxygen from the fuel, partially dissolving the Analysis of fuel composition (solid solution composed of

fuel through reduction. This interaction between cladding and fuel [U0,86Zr0,12Fe0,005Cr0,001 Nd0,006Pu0,004Ce0,004]O2,08-2,11) after begins at 1,273 K, when the materials are in contact, which is FPT1 showed that the fuel had, in fact, reacted with the oxidized generally the case for an irradiated fuel rod. At 2,023 K, liquefac- cladding, as well as with structural materials (such as iron in tion of the cladding accelerates the interaction process which an oxidized state), and that even the fuel itself had oxidized. To ultimately leads to loss of fuel integrity at a temperature of determine the liquefaction temperature of this material, and

approximately 1,000 K, below the UO2 melt point (3,120 K). the relative influence of the various effects (irradiation, cladding In a high-oxidation environment, the phenomenology is con- and structural materials, fuel oxidation) on this temperature, siderably different. Cladding exposure to this type of atmosphere the MEPHISTA database naturally appeared as an appropriate reduces the interaction process described previously, since the tool. cladding is quickly oxidized by the ambient environment, and

interaction between the oxidized cladding (ZrO2) and the fuel All the experiments conducted on irradiated uranium dioxide

(UO2) only begins at very high temperature (2,800 K, according show that fission products have only a slight effect on its liq-

to the UO2-ZrO2 phase diagram), leading to "late-phase" fuel uefaction for burnup fractions up to 5 at % [Bates, 1970; liquefaction. Yamanouchi et al., 1988]. The presence of fission products therefore cannot be put forward to explain why the fuel collapse This understanding of fuel degradation phenomena as a function temperature was lower in FPT0 and FPT1. of the ambient atmospheric conditions prevailed for a long time, until the conclusions established after the first two tests conducted in the PHÉBUS-FP Program (FPT0 and FPT1), dedicated to studying core meltdown in an accident situation on a PWR

108 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.4

T = 2,573 K T = 2,673 K 1 O 0 1 O 0 G G 0.9 0.1 0.9 0.1

0.8 0.2 0.8 0.2

0.7 0.3 0.7 0.3

0.6 0.4 0.6 0.4

0.5 0.5 0.5 0.5

0.4 0.6 0.4 0.6

0.3 0.7 0.3 0.7

0.2 0.8 0.2 L 0.8 0.1 0.9 0.1 0.9

0 F 1 0 F 1 O2U1 O2Zr1 O2U1 O2Zr1 1 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 0 1 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.10

Molar fraction Molar fraction P = 1 atm P = 1 atm

Figure 7 UO2-ZrO2-O phase diagram at Ptot = 0.1 MPa at 2,573 K (left) and 2,673 K (right).

Composition TL(K) TS(K) Composition TL(K) TS(K)

(U0.88Zr0.12) O2.000 3,080 3,020 (U0.88Zr0.12) O2.08-2.11 2,980 - 2,960 2,760 - 2,660

(U0.87Zr0.12Fe0.01) O2.08- (U0.87Zr0.12Fe0.01) O2.00 3,060 2,860 2,960 - 2,920 2,560 - 2,460 2.11

Table 2 Liquidus (TL) and solidus (TS) temperatures, calculated by MEPHISTA for compositions representative of those measured after FPT1.

Thermodynamic calculations (Table 2) show that fuel oxidation evaluations reveal the role that fuel oxidation could play in fuel alone significantly lowers the temperature at which it begins liquefaction under core meltdown accident conditions, and the to liquefy (solidus temperature). This effect could also be called need to introduce this mechanism in computer codes that on to explain the fuel failure temperatures measured in the simulate core degradation to improve quantification of the VERCORS tests, conducted in an oxidizing atmosphere [Pontillon material mass that flows to the bottom of the reactor as a et al., 2005]. function of time, and thereby assess the risks of ex-vessel progression. In the calculations, occurrence of the liquid phase for the above composition is related to modeling of the various phases (liquid Behavior of fission products in TRISO fuels planned [L], overstoichiometric solution [F] and gas [G]) at high tem- for use on high-temperature reactors perature in the U-O binary system (Figure 6) and extrapolation The very-high temperature reactor (VHTR) is one of the tech- th of this model to the composition domain bounded by UO2, nological options being studied in the development of 4

ZrO2 and O in the U-O-Zr ternary system (Figure 7). Modeling generation nuclear reactors. of the U-O binary system takes into account the oxygen poten- This type of reactor uses thermal neutrons, is cooled with tial of the UO2+x overstoichiometric solution [Chevalier et al., gaseous helium, and has a graphite moderator. The fuel, which 2004], and recent measurements performed by [Manara et al., reaches temperatures ranging from 1,273 to 1,473 K in normal

2005] on liquefaction of UO2+x solid solutions. These two operation, is based on a special technology employing UO2 sources of information, which unequivocally set the Gibbs particles (called "TRISO particles") less than one millimeter in energies of the various phases at high temperature, support the diameter, coated with different layers of material (Figure 8), validity of the described thermodynamic calculations. These each having its own specific function.

IRSN - 2008 Scientific and Technical Report 109 2.4

T/K - Temperature 3,600

3,200 2,800

2,400 FCC_B1 High-density pyrocarbon 2,000 layer (40 µm) BCT_X 1,600 1,200 SiC layer (35 µm) U1(BCC_A2)-> C3U2(S)-> U1(TFT)-> 800 <-C1(GRA_HEX_A9) <-C1U1(FCC_B1) High-density pyrocarbon 400 U1(ORT_A20)-> layer (40 µm) 0 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 Porous pyrocarbon layer (95 µm) C U X – Molar fraction Fuel kernel (500 µm) (43Car) (52Mal) (61Wit) (66Sea) (43Sno) (59Bro) (63Wit) (69Ben) (52Mal) (60Blu) (63Wit) MEPHISTA – (52Chi) (60Blu) (65Gui)

Figure 8 Cross-sectional view of a TRISO particle. Figure 9 Calculated C-U phase diagram (MEPHISTA shown as solid line) compared with experimental points.

In particular, the containment layers (two high-density pyro- and only a few measurements are available. Estimates were made carbon layers and a silicon carbide layer) are separated from by [Lindemer and Nordwall, 1974], for particles irradiated to the fuel kernel by a porous layer (called the buffer), designed 6 at %, that indicate values between –480 and –380 kJ/mol at to contain gases released by the kernel (fission products, carbon 1,450 K, and between –500 and –380 kJ/mol at 1,873 K. monoxide and carbon dioxide resulting from interaction between

the oxygen released by the UO2 kernel during fission and carbon Thermodynamic calculation of CO and CO2 generation depends from the buffer) and to limit the expansion of the outer layers on modeling of the U-C-O ternary system. In MEPHISTA, this caused by irradiation of the fuel kernel. model is based on modeling of both the U-O binary system (Figure 3) and U-C binary system (Figure 9), as well as model- The containment function of the different layers is crucial, given ing of the ternary phases, based on a bibliographic review [Potter, the burnup fractions targeted by this reactor system (up to 1972]. MEPHISTA calculates the CO pressures measured above 15 at %). In this context, it is essential to demonstrate that the the different composition domains of the U-C-O ternary system containment layers are able to withstand the mechanical load [Gossé et al., 2006] without any preliminary adjustments, exerted by gases (carbon monoxide and gaseous fission prod- thereby correctly validating the thermodynamic model. ucts), and that the fission products (FP) themselves are retained inside the TRISO particles. Figure 10 shows the change in carbon monoxide and carbon

dioxide (CO and CO2) pressures in the free space of the TRISO To accurately assess these functions, the kinetic aspects of particle buffer at 1,450 K and 1,873 K as a function of the interaction between the chemical elements must be examined oxygen potential. The temperatures selected correspond to [Barrachin et al., 2008], which cannot be done using thermo- current VHTR design-basis data for normal operating conditions dynamics. The physical state and the possible chemical form of and the maximum admissible temperature in accident situations. the various fission products can, however, be assessed through Calculations show that the pressure leading to failure of the thermodynamics, as well as the amount of carbon monoxide SiC layer (estimated at 100 MPa) is obtained for an oxygen and carbon dioxide generated at equilibrium (in a conservative potential between –330 and –325 kJ/mol. approach).

The chemical form of fission products and generation of CO and This value, measured on UO2 fuel of a PWR reactor for a bur-

CO2 gases depend on the oxygen potential and temperature, nup fraction of about 9 at % [Walker et al., 2005], should be among other parameters. Oxygen potential is difficult to evaluate reached by VHTR fuel at higher burnup, given the respective

110 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.4

Pressure (MPa) 1,000 Fraction 1

100 0.8

10 0.6

1 0.4

0.2 0.1

0 0.01 -500 -475 -450 -425 -400 -375 -350 -325 -300 -600 -550 -500 -450 -400 -350 -300 Oxygen potential (kJ/mol) Oxygen potential (kJ/mol)

CO (1,450 K) C60 Cs CO2 (1,450 K) Cs Mo0 CO (1,873 K) 2 4

CO2 (1,873 K)

Figure 10 Change in CO and CO2 pressure at thermodynamic Figure 11 Change in the cesium distribution in the different phases at equilibrium in the buffer of a TRISO particle at 1,450 K and 1,450 K as a function of oxygen potential. 1,873 K as a function of oxygen potential. differences in 235U enrichment. Assuming that fission gases (Xe these stoichiometric compounds come from measurements and Kr) are completely released from the kernel, their relative taken between 670 K and 1,070 K [Salzano and Aronson, 1965; contribution to total internal pressure may be significant for Salzano and Aronson, 1966]. moderate burnup (around 5 at %), without jeopardizing the mechanical strength of the particle. At higher temperatures (corresponding to an accident situation), decomposition of these compounds could result in cesium The contribution of the other volatile fission products to total release from the buffer. Based on the thermodynamic data and pressure is much smaller. Nonetheless, examination of TRISO current understanding of the type of chemical bonds in these particles after irradiation [Minato et al., 1994] showed that compounds, an approach was developed by [Salzano and Aronson, while the fissile kernel generally retained rare earth elements 1967] to predict the conditions in which cesium carbides may (La, Nd, etc.) and part of the alkaline earth elements (Ba, Sr), be formed or broken down. certain fission products were capable of diffusing to the buffer and becoming trapped (Cs), even migrating outside the particle Using this approach shows that at a temperature of 1,873 K, (for certain metal fission products such as Ag), with no explana- cesium should not bind with carbon. tion found for these migration mechanisms. It therefore should contribute to gaseous release from the TRISO The MEPHISTA database can serve to study the chemical form particle. This point is currently being studied within the context of cesium in the TRISO particle, which has a crucial influence of the European RAPHAEL Program [Phélip et al., 2006]. on cesium release in normal operation and in accident conditions. Thermodynamic calculations show that in the presence of carbon (from the buffer), for an oxygen potential less than

–385 kJ/mol, the formation of cesium carbides CnCs (n = 8,10,24,36,48,60) may occur at 1,450 K, consequently trapping cesium in the buffer (Figure 11).

The stability of these compounds, however, was only studied at relatively low temperatures. The thermodynamic data from the MEPHISTA database used to build the Gibbs energies of

IRSN - 2008 Scientific and Technical Report 111 2.4

Conclusions and outlook Lastly, researchers envisage using the MEPHISTA database to support the development of thermodynamic models Initiated by IRSN, development of the MEPHISTA thermo- adapted to the MFPR (Module for Fission Product Release) dynamic database began eight years ago, in cooperation computer code, covering the behavior of fission products with Thermodata, INPG and CNRS. Today it contains the and the microstructure evolution in irradiated fuels, devel- model of a relatively vast chemical system that can be used oped at IRSN in cooperation with IBRAE (Russia). to study a large number of thermodynamic issues involving either oxide or carbide irradiated fuels. Future modeling Acknowledgements work aims to explore these subjects in two main directions: The author would like to thank B. Cheynet (Thermodata) extension of the MEPHISTA database to include other ele- and E. Fischer (INPG) for the thermodynamic modeling ments, such as fuel doping elements, to obtain a more work they have conducted over the years, including the precise description of how oxygen potential changes as a MEPHISTA database and NUCLEA database, dedicated to function of irradiation, a key thermodynamic parameter for corium thermodynamics. He also thanks his colleagues at determining the chemical form of fission products; IRSN, R. Dubourg and M. Kissane, for their cooperation on analysis and incorporation of experimental data resulting work involving the behavior of fission products in TRISO from various experimental programs in progress (such as fuel, and F. Gupta for contributing to this overview. ISTC PRECOS).

112 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.4

References M. Barrachin (1993). Contribution à l’étude de l’ordre local dans le système nickel-vanadium par diffusion diffuse de neutrons, doctorat de l’université Paris XI-Orsay. M. Barrachin, F. Defoort (2004). Thermophysical properties of in-vessel corium: MASCA programme related results, Proceedings of the OECD MASCA Seminar, Aix-en-Provence (France), 10-11 June. M. Barrachin, P.-Y. Chevalier, B. Cheynet, E. Fischer (2007). New modelling of the U-O-Zr phase diagram in the hyper-stoichiometric region and consequences for the fuel rod liquefaction in oxidising conditions, J. Nucl. Mater. 375, 397.

M. Barrachin, R. Dubourg, M. Kissane, V. Ozrin (2008). Progress in understanding fission-product behaviour in TRISO-coated UO2 fuel particles, European Material Reseach Society (EMRS 2008), Strasbourg (France), May. J.-L. Bates (1970). Melting Point of Irradiated Uranium Dioxide, J. Nucl. Mater. 36, 234. S.-V. Bechta, E.-V. Krushinov, V.-I. Almjashev, S.-A. Vitol, L.-P. Mezentseva, Yu.-B. Petrov, D.-B. Lopukh, V.-B. Khabensky, M. Barrachin, S. Hellmann, K. Froment, M. Fischer, W. Tromm, D. Bottomley, F. Defoort, V.-V. Gusarov (2006). Phase diagram of the ZrO2-FeO system, J. Nucl. Mater. 348, 114. S.-V. Bechta, E.-V. Krushinov, V.-I. Almjashev, S.-A. Vitol, L.-P. Mezentseva, Yu.-B. Petrov, D.-B. Lopukh, V.-B. Khabensky, M. Barrachin, S. Hellmann, K. Froment, M. Fischer, W. Tromm, D. Bottomley, F. Defoort, V.-V. Gusarov (2007). Phase diagram of the UO2-FeO1+x system, J. Nucl. Mater. 362, 46. P.-Y. Chevalier, E. Fischer (2001). Thermodynamic modeling of the C-U and B-U binary systems, J. Nucl. Mater. 288, 100. P.-Y. Chevalier, B. Cheynet, E. Fischer (2004). Progress in the thermodynamic modeling of the U-O-Zr ternary system, Calphad 28, 15. B. Cheynet, P. Chaud, P.-Y. Chevalier, E. Fisher, P. Masson, M. Mignanelli (2004). NUCLEA Propriétés dynamiques et équilibres de phases dans les systèmes d’intérêt nucléaire. Journal de Physique IV, 113, 61. B. Cheynet, E. Fischer (2006). MEPHISTA: A Thermodynamic database for new generation nuclear fuels, http://hal.archives-ouvertes.fr/hal-00222025/fr/. B. Clément, N. Hanniet-Girault, G. Repetto, D. Jacquemain, A.-V. Jones, M.-P. Kissane, P. von der Hardt (2003). LWR severe accident simulation: synthesis of the results and interpretation of the first PHÉBUS-FP experiment FPT0. Nucl. Eng. Des. 225, 5. E.-HP. Cordfunke, R.-JM. Konings (1990). Thermochemical data for reactor materials and fission products, North Holland. A. Debremaecker, M. Barrachin, F. Jacq, F. Defoort, M. Mignanelli, P.-Y. Chevalier, B. Cheynet, S. Hellmann, F. Funke, C. Journeau, P. Piluso, S. Marguet, Z. Hózer, V. Vrtilkova, L. Belovsky, L. Sannen, M. Verwerft, P.-H. Duvigneaud, K. Mwamba, H. Bouchama, C. Ronneau (2003). European thermodynamical database for in and ex-eessel applications, FISA 2003, Luxembourg (Luxembourg), 10-13 November. A. Finel (1987). Contribution à l’étude des effets d’ordre dans le cadre du modèle d’Ising : états de base et diagrammes de phase, doctorat d’État de l’université Pierre et Marie Curie. S. Gossé, C. Guéneau, C. Chatillon, S. Chatain (2006). Critical review of carbon monoxide Pressure measurements in the uranium-carbon-oxygen ternary system, J. Nucl. Mater. 352, 13. F. Gupta (2008). Étude du comportement du produit de fission césium dans le dioxyde d’uranium par méthode ab initio, doctorat de l’université Paris XI-Orsay. M. Hansen, K. Anderko (1991). Constitution of binary alloys, Genium Publishing Corporation, New York. F.-C. Iglesias, B.-J. Lewis, P.-J. Reid, P. Elder (1999). Fission product release mechanisms during reactor accident conditions, J. Nucl. Mater. 270, 21. D. Jacquemain, S. Bourdon, M. Barrachin, A. Debremaecker (2000). FPT1 Final Report, IP/00/479. L. Kaufman, H. Bernstein (1970). Computer calculation of phase diagrams with special reference to refractory metals, Academic Press, New York. H. Kleykamp (1985). The chemical state of fission products in oxide fuels, J. Nucl. Mater. 131, 221.

T.-B. Lindemer, H.-J. de Nordwall (1974). An analysis of chemical failure of coated UO2 and other oxide fuels in the high temperature gas-cooled reactor, Technical Report ORNL-4926, Oak Ridge National Laboratory. D. Manara, C. Ronchi, M. Sheindlin, M. Lewis, M. Brykin (2005). Melting of stoichiometric and hyperstoichiometric uranium dioxide, J. Nucl. Mater. 342, 148.

K. Minato, T. Ogawa, K. Fukuda, M. Shimizu, Y. Tayama, I. Takahashi (1994). Fission product behavior in Triso-coated UO2 fuel particles, J. Nucl. Mater. 208, 266. M. Phélip, F. Charollais, S. Shihab, K. Bakker, P. Guillermier, P. Obry, T. Abram, M.A. Fütterer, E. Toscano, H. Werner, M. Kissane (2006). High-temperature reactor fuel technology in the european projects HTR-F1 and RAPHAEL. Proceedings of the 3rd International Topical Meeting on High temperature reactor technology, 1-4 Oct., Johannesburg, South Africa. Y. Pontillon, P.-P. Malgouyres, G. Ducros, G. Nicaise, R. Dubourg, M. Kissane, M. Baichi (2005). Lessons learnt from VERCORS tests. Study of the active role played by UO2-ZrO2-FP interactions on irradiated fuel collapse temperature, J. Nucl. Mater. 344, 265. P.-E. Potter (1972). The uranium-plutonium-carbon-oxygen systems-the ternary systems uranium-carbon-oxygen and plutonium-carbon-oxygen, and the quaternary system uranium-Plutonium-carbon-oxygen, J. Nucl. Mater. 42, 1. I. Prigogine (1950). Thermodynamique chimique, R. Defay (Ed.), Dunod. O. Redlich, A.-T. Kister (1948). Algebraic representation of thermodynamic properties and the classification of solutions, Ind. Eng. Chem. 40, 345. C. Ronchi, M. Sheindlin (2002). Laser-pulse melting of nuclear refractory ceramics, Int. J. of Thermophysics 23(1), 293. M. Salay, F. Fichot (2005). Modelling of Metal-Oxide Corium Stratification in the Lower Plenum of a Reactor Vessel, International Topical Meeting on Nuclear Thermal-Hydraulics (NURETH-11), Avignon (France). F.-J. Salzano, S. Aronson (1965). Thermodynamic Properties of the Cesium-Graphite Lamellar Compounds, J. Chem. Phys. 43, 149. F.-J. Salzano, S. Aronson (1966). Stability of Phases in the Cesium-Graphite System, J. Chem. Phys. 45, 2, 221. F.-J. Salzano, S. Aronson (1967). The Compatibility of Graphite with Cesium, Nucl. Eng. Des. 28, 51. B. Sundman, J. Agren (1981). A Regular Solution Model for Phases with Several Components and Sublattices Suitable for Computer Applications, J. Phys. Chem. Solids 42, 297.

C.-T. Walker, V.-V. Rondinella, D. Papaioannou, S. Van Winckel, W. Goll, R. Manzel, (2005). On the Oxidation State of UO2 Nuclear Fuel at a Burn-up of Around 100 MWd/kgHM, J. Nucl. Mater. 192.

B.-T.-M. Willis (1964). Structures of UO2, UO2+x and U4O9 by Neutron Diffraction, J. Phys. France 25, 431. S. Yamanouchi, T. Tachibana, K. Tsukui, M. Oguma (1988). J. Nucl. Sci. Tech. 25(6), 528.

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PROTECTING NUCLEAR FACILITIES AGAINST AIRCRAFT 2.5 CRASHES: impact studies on deformable missiles conducted by VTT in Finland

François TARALLO Studies examining the impact of an Finnish safety authority (STUK) asked the Civil Engineering aircraft crash on the safety of a nuclear public research institute VTT (Valtion and Structural Analysis Unit facility are carried out by facility operators Teknillinen Tutkimuskeskus), to build a facil- and are assessed by IRSN. They focus ity for testing missile impact tests on mainly on the mechanical behavior of civil targets (metal plates and concrete slabs). engineering structures subjected to In 2006, IRSN joined this program, in which impacts, and seek to assess the the first phase being completed in early 2009, consequences in terms of facility safety and the second phase scheduled from 2009 requirements. to 2011.

Up until the 1990's, these studies primar- The purpose of the program is to conduct ily covered specific hazards involving mod- test analyses using conventional materials erate degrees of energy (a light touring resistance methods in order to improve airplane, military aircraft) and used empir- understanding of the physical phenomena ical methods to study structural behavior observed and enhance numerical simulation (slab perforation). capability in this area. The main test parameters have therefore been scaled appropri- In the last few years, the definition of the ately for the study of aircraft impact on hazards to be considered has changed (more reinforced concrete structures. extensive and powerful impacts), and the computing power of software and servers The first test program phase covers the two has advanced considerably, making it essential subjects in assessment of civil engi- feasible to take into account interaction neering structural behavior in reaction to between the missile and the target in a impact: first, the missile deformation mech- transient state through fast dynamic anisms and the corresponding impact loads, analysis. and second, the mechanical behavior of the "target" structure subjected to these impact Nonetheless, these new computing capa- loads. bilities must be accompanied by simple and easy-to-model tests to provide an in-depth The first subject was studied through understanding of the phenomena at work impact tests on deformable metal tubes on and to validate the simulation tools. "rigid" steel plates instrumented to measure To investigate the impact of commercial contact loads, as close as possible to the aircraft on an EPR reactor, in 2004, the impacted surface. These tests aimed to col-

114 2008 Scientific and Technical Report - IRSN newsflashnewsflashnewsflashnewsflashnewsflashnewsflash

lect data on a point that is essential in assessments in numerical simulation of struc- ing a deformable missile impact on a target, tural elements subjected to loads leading i.e., the contact load exchanged between the to deformation of steel and concrete mate- missile and the target throughout the dura- rials in the non-linear domain, even up to tion of the impact. failure. Numerical simulations using a computer code to resolve fast-dynamic mechan- Since the publication of the method ical problems (LS-DYNA) had been described by Dr. Riera(1) in 1968, this load conducted previously and gave satisfactory is generally assessed through calculations, results in the case of slabs subjected to taking into consideration the distribution bending. of mass and crash load along the centerline of the missile. Part of the VTT tests vali- Based on an experimental program to dated the practical choice of these distribu- qualify the VTT tests concrete specimens, tions and identified limits in applying the IRSN would like to see the development method. of a behavior law specific to concrete sub- The initial results suggest that, under com- jected to high dynamic loads, to be used parable experimental conditions, the mis- in the LS-DYNA and the CAST3M finite- siles filled with water would convey to the element codes. target a momentum greater than that con- Moreover, simulations conducted using a veyed by missiles that do not contain water, simplified slab model can be used to quan- which would convey momentum compa- tify the macroscopic parameters that drive rable to their momentum drag before the dynamic bending behavior, i.e. slab stiffness, impact. which determines slab vibration frequency and the plastic moment of reinforced con- Consequently, the Riera method would crete sections that determine its final dis- not be strictly applicable to the case of placement. missiles containing a significant amount of liquid. This phenomena, which can Continued pursuit of work to define and significantly affect safety analyses on certain interpret these tests should lead to a meth- external hazards, must be confirmed by od for simulating the behavior of civil engi- dedicated tests, which will be performed in neering structures subjected to impact. This the second phase of the experimental method will be applicable to actual struc- program. tures and will serve to support safety assessments conducted by IRSN. The second topic was studied in impact tests using the same metal tubes on square reinforced concrete slabs with 2-meter sides, from 15 cm to 25 cm thick. One of the main objectives of these tests was to predict displacement and structural damage in the reinforced concrete structures loaded beyond their yield strength.

These tests revealed the main operating (1) J. Riera (1968). On the stress analysis of structures modes of reinforced concrete (bending and subjected to aircraft impact forces. Nuclear engineering and design, 8:415-426. shear strength) and contributed to improve-

IRSN - 2008 Scientific and Technical Report 115 2.6 SARNET: major achievements and outlook

Thierry ALBIOL Department of Studies and Experimental Research on Chemistry and Fires Jean-Pierre VAN DORSSELAERE Major Accident Prevention Division Bernard CHAUMONT Research Program Division Tim HASTE Paul Scherrer Institut Christophe JOURNEAU French Atomic Energy Commission (CEA) Leonhard MEYER Forschungzentrum Karlsruhe GmbH Bal Raj SEHGAL AlbaNova University Center Bernd SCHWINGES, David BERAHA Gesellschaft für Anlagen und Reaktorsicherheit GmbH Alessandro ANNUNZIATO, Roland ZEYEN European Commission

Fifty-one organizations have pooled their research capacity in SARNET (Severe Accident Research NETwork of Excellence), to work on what are considered as the most important topics for improving safety in existing and future nuclear power plants in terms of severe accident conditions. Co-funded by the European Commission (EC) through the 6th Framework Program, this program was established to optimize the use of available resources and build sustainable research groups in the European Union. SARNET has set out to reverse the fragmented state of the various existing national R&D efforts by defining joint research programs and developing shared methodologies and computer tools for safety assessment. SARNET includes almost all organizations partici- pating in severe accident research in Europe and Canada.

To achieve these goals, all organizations belonging to the SARNET harmonize and re-orient existing research programs and define network contribute to a joint program of activities with the new programs; following objectives: analyze experiment results generated by research programs implement an Advanced Communication Tool to access all to develop a shared view of the corresponding phenomena; project information, promote information exchange, and man- develop the ASTEC European reference code (an "integral" age documentation; code used to predict the behavior of nuclear power plants in a

116 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.6

postulated severe accident situation), taking advantage of Severe accident research programs were (and still are) initiated knowledge generated on physical models within SARNET; at the national level. Cooperation agreements are often built develop scientific databases that store all research program around national programs, but on a case-by-case basis. Faced results in a shared format (DATANET); with an inevitable need to reduce national budgets in this area, develop a shared methodology for conducting probabilistic it was necessary to coordinate research more effectively so as safety assessments on nuclear power plants; to optimize the human and experimental resources employed prepare training courses and a textbook on severe accidents towards the improvement of safety on existing and future aimed at students and researchers; nuclear power plants. promote mobility of personnel between the different European organizations. Consequently, in April 2004, 51 organizations, including technical safety organizations (TSOs), industry, research institutes This article presents major achievements accomplished in the and universities, decided to take the opportunity offered by the last four and a half years of networking new knowledge, along EC in the context of the 6th Framework Program to group their with improvements on the ASTEC European reference code, severe accident research capabilities together in a lasting struc- dissemination of results, and integration of research programs ture that became SARNET (Severe Accident Research NETwork conducted by the different partners. of Excellence). These organizations come from 18 Member States of the European Union (Austria, Belgium, Bulgaria, the Czech Following this initial period (2004-2008), co-funded by the EC, Republic, Finland, France, Germany, Greece, Hungary, Italy, another four-year contract for the coming years is being nego- Lithuania, the Netherlands, Romania, Slovakia, Slovenia, Spain, tiated with the EC as part of the 7th Framework Program. During Sweden, and the United Kingdom), as well as Switzerland, Canada this period, the network's activities will focus mainly on prob- and joint EU research centers (Figure 1). lems considered as top-priority during the first period; experimental activities will be included directly in collaborative work and the network will evolve towards complete self-sufficiency. Sweden The principles of this development plan are presented in the Canada Finland

N last section of this article. S Severe O E

A Accident S R Research United Kingdom NET NETwork of excellence Lithuania 800 to 900 people/month every year € SARNET background and goals An annual budget of about 24 million Czech Republic € Nether- (of which 6.8 million is financed by the lands Slovakia Germany European Community for 4-5 years) Belgium 20 countries (Europe plus Canada) Luxembourg • 51 organizations Nuclear power plants operated in Europe are designed accord- Switzerland Austria Hungary • 19 research centers Romania • 10 universities ing to the principles of defense-in-depth. They feature robust France Italy Slovenia • 11 private corporations Bulgaria engineering and containment systems conceived to protect the • 4 locations Spain • 7 safety authorities or population from radioactive release as depicted in a complete technical safety organizations Greece • Over 230 researchers series of postulated accident scenarios. • Nearly 20 doctoral students

Nonetheless, in certain highly improbable circumstances, severe Figure 1 General features of SARNET. accident conditions may cause core meltdown, leading to plant damage and dispersion of radioactive substances in the environment, thereby creating a risk for public health.

In 2004, several studies had already been conducted on severe accidents in water reactors, particularly through numerous European efforts engaged in the European Commission's 4th and 5th Framework Programs. In spite of these projects, there were still several topics requiring research work to reduce what was considered as significant uncertainty, and consolidate severe accident management plans [Magallon et al., 2005].

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The network benefits from the complementarity that exists are composed of executives from the "end-user" organizations, between the various partners (experts in fission product mate- including industry, electrical power operators, and regulatory rials and chemistry, small-scale/large-scale testing, simulated bodies (who are not necessarily members of SARNET), appoint- and real materials, experimenters, model developers, code devel- ed by the Governing Board. opers) and reinforces their strengths through synergy [Micaelli et al., 2005]. The Management Team, operating on behalf of the Governing Board, is in charge of the daily administration of SARNET. SARNET aims to achieve broad general goals: In keeping with the Joint Program of Activities (see section to counter the fragmented state that exists between the below), the Management Team consists of the SARNET various R&D organizations, in particular by defining joint research Coordinator (the team leader), the Advanced Communication programs, and by developing and validating IT tools; Tool Leader, the ASTEC Coordinator, the PSA2 Leader, the to harmonize methodologies applied in risk assessment and Experimental Database Leader, the Severe Accident Research improve Level 2 probabilistic safety assessment (PSA) tools; Priorities Leader, three Scientific Coordinators (for corium, to disseminate knowledge to newcomers in the European containment, and source term research), the Excellence Spreading Union more efficiently and promote their involvement in the Coordinator, and a Secretary. definition and accomplishment of research programs; to bring together preeminent scientists specialized in severe The Management Team monitors progress made in the various accident research to establish world leadership in IT tools used activities, works with the relevant project leaders to examine in severe accident risk assessment; what action can be taken to overcome any problems, reviews to train the next generation of severe accident experts. new projects, promotes cooperation within the network and with external organizations (OECD, ISTC, IAEA, etc.), makes This article presents major achievements accomplished in the proposals to the Governing Board to update the program of last four and a half years of networking new knowledge, as well activities, and circulates information inside and outside the as improvements on the ASTEC European reference code for network, particularly by regularly organizing specialized confer- severe accidents, dissemination of results, and integration of ences and seminars via the SARNET public website. The research programs conducted by the different partners. Coordinator operates under the supervision of the Governing Board and reports to this authority. He/she presents technical and financial reports to the Governing Board and implements Organization its decisions, in particular by updating the program of activities. The Coordinator is also responsible for relations with the EC The SARNET network is organized in a two-level structure. The which, once a year, organizes a review on the project's state of first level consists of a Governing Board, in charge of strategic progress, carried out by a panel of independent experts. decisions, supported by an Advisory Committee. The second level is occupied by the Management Team, responsible for managing the network on an everyday basis. Joint Program of Activities

The Governing Board examines progress made by the network, To meet SARNET goals, a Joint Program of Activities (JPA) is especially with regards to gradual integration, and makes recom- established and updated every year. All members of the SARNET mendations on future orientations. It decides how the EC finan- network contribute to the JPA, divided into 20 internal projects cial contribution will be apportioned and approves the research (Figure 2) that can be grouped into three categories: activity program. The Governing Board is made up of one delegate from each party to the agreement and an EC representative, who participates as an observer. In general, the members are ranking executives in their home organization and may call up resources from their organization to execute SARNET activities. The Advisory Committee advises the Governing Board on strategic orientations for SARNET research activities. Its members

118 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.6

Integrating activities Joint research activities Spreading knowledge

WP 1: ACT WP 2: ASTEC - USTIA WP 9,10,11: CORIUM WP 17: ET Advanced ASTEC Users Support • Early-phase core degradation Education and Training Communication Tool and Training, • Late-phase core degradation Integration and • Ex-vessel corium recovery WP 18: BOOK Adaptation WP 6: IED Book on severe accident phenomenology WP 12,13: CONTAINMENT Implementation of an WP 3: ASTEC - PHYMA Experimental Database • Hydrogen behavior WP 19: MOB ASTEC PHYsical Model • Fast interaction in containment Mobility program Assessment WP 7: SARP WP 14,15,16: SOURCE TERM Severe Accident WP 4: ASTEC - RAB • FP Release and Transport Research Priorities • Impact of Aerosol Behavior on Source WP 20: MANAGEMENT ASTEC Reactor Term Application • Impact of Containment Chemistry on WP 8: IA Benchmarking Source Term Integration Assessment WP 5: ASTEC - PSA2 Level 2 PSA methodology and advanced tools

Figure 2 SARNET Joint Program of Activities.

integration activities to reinforce relationships between contact and communication between partners (interactive organizations; and collaborative services); joint research activities to advance knowledge; coordination of action and programs (cooperative network activities to spread excellence and knowledge. management); links to projects carried out by satellite communities (R&D projects, related sites). Main accomplishments The ACT platform was built based on the MS Sharepoint Portal After four and a half years of operation, the network can proud- Server. Access is provided through web browsers, via any Internet ly claim the achievements described below. connection. Access to this tool has been granted to about 250 SARNET users. ACT is accessed intensively and efficiently, with Integration activities 500 to 1,000 connections per month and 8,000 items in storage The integration aspect of the program is given utmost impor- (Figure 3). tance and projects in this area are considered as key items in A public website (www.sar-net.org) also provides essential the JPA. information on SARNET. The Advanced Communication Tool (ACT) was implemented to promote communication between project partners and ensure DATANET, the SARNET experiment database network, was devel- documentation management. ACT is a key step towards meet- oped to preserve, exchange, and process experimental data on ing SARNET goals: the current technology based on portals severe accidents, including any related documentation. Data establishes a unified system for efficient cooperation across comes from existing experimental data that SARNET partners the network by providing: wish to share within the network, and any new data generated access to documents and computer codes, search functional- under the SARNET program. DATANET is based on the STRESA ity, and publication (the knowledge base concept); tool developed by JRC Ispra and consists of a network with

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Advanced Communication Tool (ACT) status • 250 users ACT • 500 to 1,000 accesses per month • 8,000 items stored (3,300 documents) • 11 topical sites

DATANET status • 9 open nodes (JRC, FZK, IRSN, CEA, AEKI, KTH, CIEMAT, FORTUM-VTT, GRS) • over 100 test results from 20 experimental sites stored in the database

An intranet portal (ACT) offers SARNET partners DATABASE read and/or write access to: FORTUM-VTT • network tools (documents, management, meeting AEKI database database info, forums, questionnaires) CEA database JRC Ispra KTH database database • experimental databases main portal • ASTEC code • links... .

GRS database CIEMAT database Public Website IRSN database FZK database www.sar-net.org

Figure 3 SARNET information systems.

several local databases (or nodes). From the central database, dation). Based on these detailed models, simplified models are it is possible to connect to other local databases, which can then established and introduced in the ASTEC code. also be accessed directly, multiplying the power and potential of this type of system. There are currently nine nodes: the Twenty-eight organizations cooperate with IRSN and GRS in central node at JRC Ispra and local nodes at FZK, IRSN, CEA, developing and assessing this code. ASTEC describes the behav- CIEMAT, FORTUM-VTT, AEKI, KTH, and GRS. Training courses ior of an entire nuclear power plant in severe accident conditions, are organized periodically at JRC Ispra. To date, the results of including management of the severe accident situation through over 100 experiments have been entered in the database procedures and engineering systems. It is used intensively by (Figure 3). IRSN to establish Level 2 PSAs for 900 and 1,300 MWe PWRs, and by GRS in consolidating Level 2 PSAs for 1,300 MWe Konvoi ASTEC, the European integral code for severe accidents [Van reactors. A growing number of partners also call on this code Dorsselaere et al., 2006], developed jointly by IRSN and GRS, to calculate complete accident scenarios. is the main vehicle of integration and contributes effectively to spreading knowledge. Moreover, most joint research activities Efficient and close cooperation between ASTEC users and devel- are related to ASTEC, since one of their basic goals is to provide opers has been implemented using ACT and the MARCUS web physical models to be incorporated in ASTEC and validate them. tool for code maintenance. Training courses on using the code Exchanging information on detailed models developed by have been organized. Three main versions of the code have been various experts through interpretation of experiment results released to partners: V1.1 mid-2004, V1.2 mid-2005 and V1.3 leads to models that are common to the different detailed codes at the end of 2006 (with update V1.3 rev2 released at the end (for example, ICARE/CATHARE and ATHLET-CD for core degra- of 2007). Code documentation has been improved considerably

120 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.6

since SARNET began. Three meetings of the ASTEC Users Club, the integration activities. Level 2 PSA is a powerful tool for the last one taking place in April 2008, gave users and develop- assessing the specific vulnerability of nuclear power plants to ers the opportunity to exchange views in straightforward and severe accidents. It assesses possible scenarios in terms of constructive discussions. frequency, loss of containment integrity, and release of radioactive substances to the environment, and quantifies the contri- CEA has advanced development of models on corium behavior bution of prevention and attenuation measures to risk reduction. in the vessel lower head and ex-vessel cooling, including valida- Various approaches are used in Europe, derived more or less tion through experiments such as LIVE (FZK), SULTAN (CEA) from those followed in the USA. The principle methods used by and ULPU (University of Santa Barbara). Other models were the various partners to establish PSAs have been described and proposed or improved after discussion between partners, such compared in a document. For several problems involving Level as silver-indium-cadmium and ruthenium release from the core, 2 PSA, a survey was taken and the answers were analyzed to boron carbide (B4C) oxidation, mass transfer of iodine and define the next steps to be taken towards harmonization. Case radiolysis oxidation in containment (see details and other studies were then carried out on specific problems, such as examples in the section below on joint research activities). New hydrogen combustion, iodine chemistry, corium cooling and improvements have been planned for the near future, such as corium/concrete interaction, significant early-phase release, the work on the mechanical resuspension of aerosols in system definition of the final stabilized states of the reactor, and the lines. ASTEC has also been used to prepare and interpret sev- Level 1 and Level 2 PSA interface methods. A certain degree of eral experiments, including QUENCH and LIVE at FZK. Preliminary harmonization has been achieved and recommendations writ- calculations of accident scenarios for boiling water reactors and ten. CANDU reactors, performed using current versions of the code, have given promising results. They provided the opportunity to A report on the current state of knowledge on dynamic reli- prioritize and specify necessary adaptations in the models. ability methods was written and the limits of classical methods that could be overcome by using these reliability methods were An advanced code validation process was carried out by the identified. The advantages of one of the possible methods (Monte partners on 65 analytical and integral experiments, often involv- Carlo dynamic simulation of event trees) were examined for a ing OECD/NEA/CSNI international standard problems. In gen- station blackout. Considerable work was carried out in a bench- eral, the results can be considered as satisfactory, and even very mark exercise (quantification of the risk of containment failure satisfactory for plant system thermohydraulics, core degradation, caused by activation of a safety system during the core degra- and the behavior of aerosols and fission products. The code has dation phase in the reactor vessel). This exercise included a been applied to a large number of severe accident scenarios for comparison between dynamic reliability methods and classical different types of nuclear power plants (PWR, Konvoi 1300, methods. Work is currently under way to determine exactly Westinghouse 1000, VVER-440, VVER-1000 and RBMK), and what is required in ASTEC to meet the needs of Level 2 PSAs has frequently been compared with other codes. Good agree- and to couple ASTEC with probabilistic tools. ment has been shown with integral codes MELCOR and MAAP4 with regards to trends and orders of magnitude in the results, The internal Severe Accident Research Priority (SARP) project and very good agreement with results of detailed codes such identified research priorities to gradually re-orient existing as ATHLET-CD and ICARE/CATHARE for core degradation. The national programs and contribute to initiation of new programs numerical robustness of ASTEC has improved greatly since in a coordinated manner, by eliminating redundancy and devel- SARNET began. oping complementarity. This activity was developed in close cooperation between participants representing technical safe- IRSN and GRS now take into consideration all requests expressed ty organizations, industry, and public institutions. In the same by users for upgrading ASTEC as they prepare the new ASTEC manner as the Phenomena Identification and Ranking Table V2 series. The first in the series, V2.0, scheduled for release in (PIRT) established within the European EURSAFE project June 2009, will apply to the EPR plant (especially the corium [Magallon et al., 2005], this activity took into account the catcher) and will include most of the detailed models from complete range of severe accident situations, from core uncov- ICARE2, IRSN's reference code for core degradation. ery to long-term corium stabilization, long-term containment Harmonizing methodology for Level 2 PSAs also comes under integrity, and environmental release of fission products. Special

IRSN - 2008 Scientific and Technical Report 121 2.6

emphasis was placed on a risk-oriented approach to truly focus This prioritization of subjects will serve to reapportion compe- on the most pertinent pending problems. Consensus was reached tencies and human resources to high-priority studies, for both in the final phase and 18 major problems were classified in four the EC 7th Framework Program and other international (OECD) categories. projects.

Six subjects were considered as requiring in-depth analysis and Joint research activities were given high priority: These activities form the foundation of SARNET R&D. In con- core cooling by reflooding and debris cooling; nection with the research priorities described above, their goal configuration of the corium pool in the reactor pit during is to tackle pending problems that have been given precedence. corium/concrete interaction, and top-cooling of corium; They have been divided into three areas: corium behavior, relocation of the molten pool in water, ex-vessel Fuel/Coolant containment integrity, and source term. Interaction (FCI); hydrogen mixing and combustion in containment; The same method was adopted for all three fields: review and impact of oxidation (Ru oxidation conditions and air penetra- selection of available, pertinent experiments with synthesis of tion for high burnup and MOX fuel elements) on the source analyses and interpretation of data provided by these experi- term; ments; review and synthesis of physical models with proposals iodine chemistry in the reactor coolant system and the for new or improved models for ASTEC. containment. Four subjects were given medium priority, implying that work Corium will be pursued as planned in the different research programs Corium behavior is a vast subject involved in more than half involved. Risk assessment has been defined through considerable the problems selected in the PIRT of the EURSAFE program advances in knowledge, but certain issues remain open: [Magallon et al., 2005]. The study of corium behavior extends hydrogen generation during reflooding and relocation of the from early-phase core degradation to late-phase corium forma- molten pool in the reactor vessel; tion in the reactor core and its relocation outside the vessel. corium cooling in the vessel lower head; vessel integrity after ex-vessel cooling; Significant efforts are currently being devoted to the develop- direct heating of containment through ejection of pressurized ment and improvement of databases on the physical properties corium after reactor vessel failure. of corium materials and corium thermodynamics. For five subjects, current knowledge was considered sufficient with regards to risk and safety, and given research in progress. Joint projects have been conducted by 22 SARNET partners These subjects were given low priority and will most likely be [Journeau et al., 2008], involving contributions to the definition closed once current projects have been completed: and interpretation of tests, benchmark exercises, and improve- corium cooling in a core-catcher with ex-vessel cooling; ment of the associated models: CCI-OECD tests on corium/ corium spreading in the reactor pit after reactor vessel failure; concrete interaction; the QUENCH-10 test on air ingress in a formation of cracks and leaks in the concrete containment fuel bundle; the QUENCH-11 test on fuel bundle degradation structure; and reflooding; QUENCH-12 test on a VVER bundle; COMET-L1 impact of aerosol behavior on the source term (in steam and L2 tests on corium/concrete interaction in 2D geometry; generator tubes and in containment cracks); LIVE tests on corium in the vessel lower head; VULCANO test impact of core reflooding on the source term. on corium/concrete interaction on real materials. Three subjects were considered as ready to be closed. Given the current state of knowledge and the low level of risk in com- Similar activities were led for the International Science and parison to other subjects involving greater risk and uncertainty, Technology Center (ISTC): PARAMETER on top flooding, METCOR no other experimental programs are required in the following on the impact of thermomechanical interaction on vessel areas: behavior, and CORPHAD on corium thermodynamics (and reactor coolant system integrity and heat distribution; enhancement of the NUCLEA database). For the international interaction between corium and ceramics in an ex-vessel Source Term project (see Source Term section below), FZK and core-catcher, cooling through bottom flooding; IRSN harmonized their test matrices on Zircaloy oxidation by

fuel/coolant interaction with a vapor explosion in a vulner- air/steam mixtures, and boron carbide (B4C) oxidation and able reactor vessel. degradation.

122 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.6

Some of the main achievements in corium research [Journeau Dynamic (CFD) codes (FLUENT, TONUS-3D and REACFLOW). et al., 2008] include the following: Results revealed certain weaknesses in the combustion models. understanding of oxidation phenomena in steam and air New tests are planned on the ENACCEF facility to study the atmospheres was improved, and oxidation correlations were interaction between hydrogen flames and spraying, and on the validated. The importance of material composition was REKO-3 facility to investigate gas ignition on hot catalytic demonstrated. Research on new cladding materials, and on surfaces. The results will be used to validate computer codes. hydrogen generation and fission product release in particular, are necessary; The main achievement in containment atmosphere mixing is

data on B4C boron carbide oxidation collected from experi- the containment spraying experiment simulation carried out ments conducted by FZK and IRSN have led to a shared inter- on a small-scale facility (TOSQAN at IRSN) and on a large scale pretation of the PHÉBUS-FPT3 integral test; (MISTRA at CEA). This work covered atmosphere depressurization launching of LIVE experiments (late-phase in-vessel degrada- and destruction of atmosphere stratification. Numerical simulation) started a new series on modeling molten pool behavior tions of these experiments were conducted using CFD codes in the reactor vessel; and lumped-parameter codes. Results demonstrated the feasibil- the various reactor vessel creep failure models were compared ity of these simulations, which could also be applied to real and a shared understanding of the OECD test OLHF-1 was containment structures within the next few years, once adequate obtained. Modeling of crack propagation is in progress; computational capacity is available. analyses on cooling heterogeneous 2D debris beds showed increased cooling capability as compared to 1D particle beds. At present, however, the resolution achieved when applying CFD These results renewed interest in this problem, from both the tools to the containment is insufficient, due to the wide range experimental and numerical point of view, including research of turbulence scales involved and the computing power required. on debris bed formation; CFD simulations of Passive Autocatalytic Recombiner (PAR) recent tests on corium/concrete interaction gave unexpected operation with simplified 2D containment models represent an results with marked ablation anisotropy for silicon-rich materials. important first step towards complete simulation of PAR/atmo- Interpretation and modeling of this behavior will be pursued; sphere interaction in actual power plants. Simulations of con- the core-catcher concepts based on spreading (EPR) and bot- densation experiments performed on the CONAN facility at tom flooding (COMET, downcomers) are under study and prog- the University of Pisa successfully assessed the different steam ress has been made in modeling; condensation models used in the CFD codes, steam condensa- validation of the NUCLEA thermodynamic and chemical tion having a significant influence on atmosphere behavior. database was extended through experiment result analyses As regards research on fast interactions, Fuel/Coolant Interaction funded by the EC and an inter-laboratory exercise assessed (FCI) was studied to further knowledge on parameters that uncertainty of these analyses using energy dispersive X-ray affect steam explosion energetics during corium relocation in spectrometry. water and to determine the risk of vessel or containment failure. Investigations focused on specific processes such as pre-mixing, Containment corium fragmentation, and particle thermal transfer. FCI research Research work on energy phenomena that could jeopardize was closely tied to Phase 1 of the OECD-SERENA program, containment integrity involves hydrogen behavior and fast which set out to assess the ability of the current generation of interactions in the containment. With regards to hydrogen FCI codes to predict reactor structural loads induced by a steam behavior, hydrogen combustion and attenuation of the associ- explosion, and to reach a consensus on the understanding of ated risk focus on the formation of combustible gas mixtures, important FCI phenomena in different reactor situations [Buck the local composition of gases, and potential combustion modes, et al., 2008]. including the reaction kinetics inside catalytic recombiners. The hydrogen distribution inside the containment is studied to assess The codes used for this purpose were mainly MC3D and IKEMIX/ the risk of high concentrations. Experimental programs on IDEMO. A certain number of experiments were performed in combustion with concentration gradients (ENACCEF at IRSN) the MISTEE and DEFOR facilities at KTH, and KROTOS at CEA and recombiner kinetics (REKO-3 at FZJ) have been conducted Cadarache, which was restarted after being moved from the and/or are under way. ENACCEF experimental results were used JRC Ispra center. These experiments aimed to further understand- for benchmarking with three different 3D Computational Fluid ing of the fragmentation and explosion of corium pools. The

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main achievement was the consensus reached on the fact that tion of iodine gas and liquid forms inside containment (also a steam explosion inside the reactor vessel would not cause extended to ruthenium). Another study involved aerosol behav- vessel failure (which closes the issue of the risk of steam explo- ior in scenarios where risk is of utmost importance: containment sion in the vessel), and the fact that an ex-vessel steam explo- bypass sequences (steam generator tube rupture, SGTR), con- sion could cause damage to the reactor pit. The level of loading tainment through-wall cracks, and thermal and mechanical in the reactor pit cannot be predicted, however, due to the wide revolatilization. The main accomplishments are described in dispersion of results. The main explanations for this dispersion detail in [Haste et al., 2006]. are uncertainty regarding the vacuum distribution in the pre- mixing zone, introducing large deviations in the initial explosion An important effort was devoted to pursuing the experimental conditions, and uncertainty regarding the explosive behavior of International Source Term Program (ISTP), initiated by IRSN, corium. These uncertainties will be investigated through exper- CEA, and EDF, with support from the EC [Clément and Zeyen, imentation and analysis in Phase 2 of the OECD-SERENA pro- 2005]. Interpretation of available AECL and RUSET test results gram and in follow-up work in SARNET. (the latter at AEKI) showed that Ru release occurs in the form of oxide, after an incubation period in which complete oxidation The second problem concerning fast interactions is Direct of the fuel and cladding occurs. The RUSET and VTT tests showed Containment Heating (DCH) after vessel failure. This includes that the forms of oxide can remain volatile in the lower corium dispersion in various reactor compartments, thermal temperature range, and then be transported to the reactor transfer processes, and chemical processes such as hydrogen containment in a stable gaseous form, a very important finding. generation and combustion. Since the consequences of DCH These tests are still in progress. Test conditions for the future are essentially related to reactor pit geometry, a database was VERDON program (CEA) [Clément and Zeyen, 2005] were built for different power plant designs, including EPR, the French defined through simulations of air ingress scenarios. These new PWR-1300, VVER-1000 and the German Konvoi reactor, through data will be used to validate ASTEC modeling. The proposed an experimental program run on the DISCO facilities (FZK). For VERONIKA experiment, in the ISTC context, on the release of the EPR and VVER-1000 plants, the DCH issue can be considered fission products from a highly irradiated VVER fuel was examined as closed due to the reactor pit geometry. Efforts will continue and SARNET proposals for a test matrix were adopted. A similar to improve the prediction capability of the MC3D CFD code on exercise was conducted on iodine chemistry through the EVAN one hand, and the COCOSYS and ASTEC codes on the other. test project, the results of which are currently being studied. Benchmark exercises revealed serious deficiencies in current FIPRED experiments (INR) provided data on self-disintegration

models: correlations on debris dispersion must be more closely of UO2 pellets. adapted to reactor pit geometry, and adequate models for hydrogen oxidation and combustion are required, since these With regards to fission product transport, IRSN interpreted the phenomena seriously jeopardize containment integrity (exper- iodine chemistry in the system based on data generated through imental results have shown that direct thermal effects alone the PHÉBUS-FP and VERCORS HT tests (the latter conducted should not jeopardize containment integrity). Based on available at CEA). In reducing conditions and with no control rod mate- experimental data, extrapolation to the reactor scale under an rials, it appears relatively clear that iodine is transported air-steam-hydrogen atmosphere needs to be established using mainly in the form of cesium iodide (and rubidium). In oxidizing combustion codes (COM3D, REACFLOW), then the modeling conditions, the question is more complicated: iodine may take parameters must be transferred to codes with DCH models (CFD the form of CsI or, in the presence of control rod materials, codes or COCOSYS), and in the final phase be adapted to other metal iodides or, if no control rod materials are present, ASTEC. conditions tend to favor HI formation. These points have not yet been confirmed. The CHIP experimental program (IRSN) Source term [Clément and Zeyen, 2005], in which VTT provides experimen- In the source term area of study, release, transport and deposi- tal support, has produced kinetic and thermodynamic data on tion of fission products were investigated, including situations iodine transport through the reactor coolant system, espe- with air ingress to the reactor vessel, i.e., in an oxidizing environ- cially for key systems such as {I-Cs-O-H}. ment. For transport and deposition, one study focused on iodine, This data will be used to improve ASTEC models and extrapolate analyzing both volatility in the reactor coolant system (for experimental results to reactor conditions. In the QUENCH-13 interaction with concomitant Ag-In-Cd release) and the propor- fuel bundle test (FZK), carried out using a powerful experimen-

124 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.6

tal facility (PSI, AEKI), robust theoretical foundations (PSI, GRS, several orders of magnitude. Mass transfer between the sump and EDF), and associated small-scale tests, the generation of Ag-In- the gas phase was studied on the SISYPHE facility: evaporation Cd aerosols after PWR control rod failure was measured for the conditions increase the rate of transfer from the liquid phase to first time, with speciation determined at defined time intervals. the gas phase and change the iodine equilibrium concentrations, This data was correlated with bundle degradation and compared as the sump iodine concentration is reduced. The well-known to that obtained from previous EMAIC (CEA) analytical tests. two-film model was extended to these conditions and entered in Calculations to interpret these test results, along with those ASTEC. Validation of a more mechanistic model is currently in from the PHÉBUS-FP tests, are currently under way. Model progress [Herranz et al., 2007]. enhancements will be carried out as required. The effect of radiation on the containment atmosphere and the Several facilities have been used to study aerosol retention in effect of metal impurities in the sump were studied in the PARIS the steam generator under SGTR conditions: PSAERO/HORIZON and Chalmers experimental programs, respectively, while (VTT), PECA/SGTR (CIEMAT) and ARTIST (PSI). On an overall Chalmers and VTT studied speciation of the oxides and iodine basis, these tests showed that the "wet" scenarios (where rup- nitroxides formed in the atmosphere through radiolysis. Organic ture occurs below the secondary water level) seem to provide iodine formation models take into account the thermal and effective particle retention and that the same "dry" scenarios radiolytic mechanisms in gaseous and liquid phases. There are may capture a fraction of the particles that ingress to the nonetheless deviations in the aqueous model, essentially with secondary side, but in much smaller amounts. The VTT tests regards to the organic sources. The data provided by the EPICUR showed that resuspension is significant for aerosol retention in tests are now being interpreted using codes such as ASTEC/ the horizontal tubes, and that it increases for sudden variations IODE, INSPECT, and LIRIC. in flow rate. All available resuspension models are being assessed through intercomparison of data (STORM at EC/JRC Ispra and The behavior of ruthenium inside containment was studied new VTT results). The VTT data will serve to develop a new experimentally and theoretically by IRSN (EPICUR tests) and empirical model. Revolatization tests in the REVAP small-scale Chalmers, along with the effects of heating fission products facility (JRC/ITU) on PHÉBUS-FP samples show that Cs revapor- on the passive autocatalytic recombiners, successfully modeled ization can be very high (approx. 95%) on flat metal substrates. by ASTEC/SOPHAEROS based on the RECI laboratory tests CsOH deposits on stainless steel showed the same behavior as (IRSN). Scaling effects are currently being studied. Lastly, a PHÉBUS-FP deposits. VTT is starting new tests on the speciation benchmark exercise using several codes was performed in the of revaporization aerosols, to complement the previously men- context of the ThAI-Iod9 integral test on containment iodine tioned CHIP program tests. chemistry, coordinated by GRS.

Aerosol retention in containment cracks can be effective, espe- Integrating research activities cially in the presence of steam (SIMIBE tests at IRSN). Independent The previous sections have demonstrated that effective integra- models or models incorporated in codes have been developed. tion of research work has been achieved through: They will be tested with respect to data taken from COLIMA collaborative work on calculations to prepare and interpret experiments (CEA), which will be performed on the experimen- experiments, for example PSI on the QUENCH tests at FZK; tal PLINIUS platform. A SARNET team will provide assistance joint experimental achievements, such as the specific facility before testing. installed and operated by VTT experimenters for CHIP experiments at IRSN; Facilities designed to study containment chemistry include collaborative definition and interpretation of experiments (sev- PHÉBUS-FP, CAIMAN, SISYPHE, Chalmers, PARIS and EPICUR eral "interpretation circles" have been created and are truly [Clément and Zeyen, 2005], as well as RTF (AECL), which recent- active); ly produced data from the P9T3 test. The Iodine Data Book, com- comparisons between computer code results; piled by Waste Management Technology, provides a critical review dissemination of codes among the partners to reach a shared of chemical data and models. CAIMAN results showed that in the vision of experimental phenomena (primarily ASTEC presence of paint, under irradiation and at high temperature, modules); organic iodine may be the predominant form of gaseous iodine; exchanges on the application of R&D results at the reactor in alkaline conditions, gaseous iodine concentrations decrease by scale;

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interlaboratory exercises for EDX analyses of prototype cori- Extension of the SARNET experience um samples by three laboratories (Cadarache in France, Karlsruhe in Germany, Rez in the Czech Republic); The SARNET contract with the EC covered a period of four and annual technical meetings in all three areas (corium behavior, a half years, from April 2004 to September 2008. In 2006, a containment integrity, and source term), in addition to a large special working group, made up of nine representatives of the number of specialist meetings. Governing Board, was created to prepare the network's con- tinuation. A positive response was obtained in the context of Spreading excellence the second request for proposals of the EC's 7th Framework The third major activity of SARNET involves spreading knowledge Program and contract negotiations are now in progress. The and excellence. The more experienced organizations began network is therefore expected to be co-funded by the EC for spreading excellence by preparing a training course on severe another four years. accident phenomenology, intended for doctoral students and In the course of this period, network activities will focus main- junior researchers, with five-day sessions organized in June 2006. ly on high-priority subjects (see section on integration activities) A five-day training course on accident progression (data, and experimental activities will be included directly in the analyses, and uncertainty) for senior nuclear safety specialists partially financed common program. The aim is to evolve towards was organized in March 2007. A third course covering both complete self-sufficiency. The basic principles defined to reach phenomenology and codes to simulate severe accident sce- this goal are explained below. narios was organized in Budapest in April 2008, where 40 to 100 people attended each course. The 41 partners from Europe, Canada, Korea, and the USA will network their research capacities in the SARNET2 project. A book on severe accident phenomenology is currently being Following through with SARNET's initial objectives, the project written. It will cover the historical aspects of water reactor was defined to rationalize the use of available resources and operating basics and safety principles, the phenomena involved build a sustainable consortium, where severe accident studies in accident progression inside the reactor vessel, containment form the basis for development of joint research programs and failure, and the release and transport of fission products. It will a shared computer tool (the integral ASTEC code). present a description of analysis tools and codes, management and limitation of the consequences of severe accidents, and In the second contract period, network organization will be environmental management. It will also provide information on similar to that of the first contract. To enhance decision-making third-generation reactors. The partners who have agreed to work efficiency, the Governing Board will be replaced by a Steering together to prepare these courses and the book are universities, Committee of 10 members in charge of strategy, supported by TSOs(1), national laboratories, and representatives of industry an Advisory Committee made up of executives from end-user who share their vast experience and talent through SARNET. organizations (for example, electricity companies).

A mobility program, allowing students and researchers to work A general assembly, composed of a representative from each in different SARNET laboratories to further their training, is of the parties to the consortium and an EC delegate, will be another aspect of spreading excellence. There have been 33 convened once a year to report on and discuss network activ- transfers funded by SARNET, for periods lasting an average of ity progress, the detailed implementation plan, and decisions three months. taken by the Steering Committee. At the second level, as for Three European Review Meetings on Severe Accident Research SARNET, a Management Team (with its Coordinator and seven (ERMSAR) were held in France, Germany, and Bulgaria, provid- Project Leaders) will be tasked with daily management of the ing forums for the severe accident research community. They network. have become major world events on the subject. The principle activities to promote integration and excellence Lastly, about 300 talks or papers have been presented by SARNET will be pursued: collecting available experimental data in a members in conferences or scientific journals. The complete list scientific database in a format common to all, integration of is given in the SARNET final summary report (open to the public) [Albiol et al., 2008]. (1) Technical safety organizations.

126 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.6

knowledge acquired through network projects in ASTEC (with By promoting collaborative work in Level 2 PSAs, SARNET special attention given to applicability to all types of nuclear has begun to create the necessary conditions for harmoniz- power plant found in Europe), dissemination of knowledge ing approaches and making Europe a leader in risk through the SARNET public website and the ACT tool, organiza- assessment methodology for severe accidents. tion of conferences and seminars, organization of teaching and training sessions, encouraging student and researcher exchang- Through a teaching and training program (organizing es. Research priorities will also be reviewed periodically. Level classes, writing a book for junior scientists), SARNET is 2 PSA activities will not be continued in the network, since a developing synergy with educational institutions so that this specific EC contract on harmonization of Level 2 PSA studies field of activity remains attractive to young people. This was established and launched in 2008 (www.asampsa2.eu). action has been reinforced by a successful mobility program that provides young students and researchers with exchange Contrary to the first contract period, joint research activities opportunities at various laboratories throughout Europe. under the second contract will include experimental programs. In keeping with EC recommendations, among the high-priority By promoting collaborative R&D work on corium behavior, issues defined in SARNET (see section on integration activities), containment integrity, and source term issues, SARNET has experimental work will be dedicated primarily to two subjects advanced towards the resolution of residual questions and considered of utmost importance, which are expected to advance provided modeling recommendations for ASTEC. Proposals towards closure: coolability of corium and debris, and corium/ have been made for various ASTEC models, for which concrete interaction. The other high-priority subjects will not implementation has been completed or is in progress. be neglected. Shared analyses of experimental results and comparisons of models and codes to achieve a shared view of Pursuing SARNET will clearly make this network a refer- the corresponding phenomena will continue through different ence in terms of severe accident research priorities, with an technical circles. Relations with end-users and other interna- impact on national programs and the corresponding bud- tional programs and organizations such as the OECD/NEA, ISTC gets. In time, all research activities in this field will be and other programs co-funded by the EC (mainly SNE-TP, coordinated by the SARNET, contributing to a more rational PHÉBUS-FP, ISTP, ASAMPSA2) will be maintained and reinforced. use of European resources. Plans also envisage a gradual extension of program activities to fourth-generation reactors. Acknowledgements Finally, the network will move towards effective financial self- We would like to extend our thanks to the following project sufficiency by creating a legal entity for this purpose. Further and sub-project leaders: M. Steinbrück (FZK), G. Repetto details are given in [Albiol et al., 2008]. (IRSN), C. Duriez (IRSN), W. M. Ma (KTH), V. Koundy (IRSN), M. Bürger (IKE), B. Spindler (CEA), H. Wilkening (JRC/IE), I. Kljenak (JSI), D. Magallon (CEA), P. Giordano Conclusion (IRSN), L. Herranz (CIEMAT). The Management Team has strongly relied on these leaders as technical relays to other The SARNET network of excellence began its activities in network members, too numerous to mention here, but to April 2004, with the ambitious but extremely important goal which the authors are grateful. Finally, and above all, the of providing the appropriate framework for lasting integra- authors thank the European Commission and its representation of European severe accident research capabilities. tive (M. Hugon, who has shown a true, profound interest in the network) for financing SARNET through the Sixth By pooling knowledge acquired in the ASTEC European Framework Program, in the area "Nuclear Fission: Safety of reference code and the DATANET database, SARNET has Existing Nuclear Installations", contract number FI6O-CT- effectively set up the necessary conditions for preserving 2004-509065. the knowledge generated by thousands of people dedicated to research, and disseminating this knowledge to a large number of end-users. The ASTEC code and DATANET are now widely used resources.

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References T. Albiol et al. (2008). Presentation of SARNET2, European Review Meeting on Severe Accident Research (ERMSAR 2008), Nesseber, Bulgaria. T. Albiol et al. (2008). SARNET Final Activity Report (Synthesis), SARNET Deliverable N° 124 (to be issued also as a EC public report with a specific number). M. Buck et al. (2008). Status of FCI Modeling and Reactor Applications: Achievements during the SARNET Project, European Review Meeting on Severe Accident Research (ERMSAR 2008), Nesseber, Bulgaria. B. Clément, R. Zeyen (2005). The PHÉBUS-FP and International Source Term Programmes, Proc. Int. Conf. on Nuclear Energy for New Europe, Bled, Slovenia. T. Haste et al. (2006). SARNET, Integrating Severe Accident Research in Europe: Key Issues in the Source Term Area, Proc. Int. Congress on Advances in Nuclear Power Plants (ICAPP ’06), Reno, Nevada, USA. L. E. Herranz, J. Fontanet, L. Cantrel (2007). Impact of Evaporative Conditions on Iodine Mass Transfer during Severe Accidents, 12th Int. Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12), Pittsburgh, Pennsylvania, USA. C. Journeau et al. (2008) European Research on the Corium Issues within the SARNET Network of Excellence, Proc. Int. Congress on Advances in Nuclear Power Plants (ICAPP ’08), Anaheim, California, USA (2008). D. Magallon et al. (2005). European Expert Network for the Reduction of Uncertainties in Severe Accident Safety Issues (EURSAFE), Nuclear Engineering and Design, vol. 235, p. 309-346. J.-C. Micaelli et al. (2005). SARNET: A European Cooperative Effort on LWR Severe Accident Research, Proc. European Nuclear Conference, Versailles, France. J.-P. Van Dorsselaere, H.-J. Allelein, K. Neu (2006). Progress and Perspectives of ASTEC Applications in the European Network SARNET, Proc. EUROSAFE Forum, Paris, France.

128 2008 Scientific and Technical Report - IRSN newsflashnewsflashnewsflashnewsflashnewsflashnewsflash

Study on iodine behavior in the reactor containment 2.7 in an accident situation: first results from the EPICUR program

Séverine GUILBERT Mechanics and Materials Experimentation Laboratory Ventilation Karine BOUCAULT, Cédric GOMEZ, Gas flow to Laurent MARTINET, Cécile PASCAL May-Pack Condenser NaI Gas Environmental and Chemical counters Experimentation Laboratory Gas flow to vessel

May-Pack filter system Condensate 60 vessel Co

Condensed steam Irradiation vessel

Condensate Steam generator reinjection pump Glovebox

Figure 1 Diagram and photograph of the EPICUR test loop.

Iodine is one of the most radiotoxic fission conditions are not sufficient to accurately products that can be released into the envi- predict the amount of volatile iodine frac- ronment during a pressurized water reactor tions that may be formed. To complete the (PWR) accident leading to core meltdown. existing database, as part of the International The amount of radioactive iodine released Source Term Program (ISTP)(1), IRSN has set subsequent to a containment leak or filtered up a research program using the EPICUR containment venting depends on the chem- irradiator to conduct tests where iodine ical form of the iodine. Changes in the con- volatization under irradiation is measured centration of volatile iodine in containment on line. depend on the balance between reactions involving the formation of iodine com- During testing, liquid-phase iodine is pounds, their fixation on surfaces, and their simulated by an iodide solution (127I), traced destruction. using radioactive iodine (131I) and placed in a 5-liter vessel simulating the contain- Iodine chemical reactions in the liquid ment (50,000 m3). The vessel is connected phase (solution present in the reactor ves- to the test loop, which includes selective sel lower head) have been studied exten- filters to trap aerosols, non-organic iodine

sively in the last thirty years and (I2) and organic iodine species (Figure 1). understanding of the mechanisms involved The volatile iodine species formed under has improved greatly, particularly through the effect of irradiation (at dose rates of tests conducted on the PHÉBUS reactor at approximately 3.5 kGy/h) and high tem- (1) The ISTP program is funded by IRSN, EDF, CEA, Cadarache. Nonetheless, the data available perature (80-120°C) were transferred by the European Commission, US-NRC, AECL, Suez- Tractebel and PSI. for high-temperature and high-dose-rate gas flow from the vessel to the filters. NaI

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Cumulative iodine fraction 10

0.1

0.01

10-3

10-4 Argon 80°C pH 5 Air 80°C pH 5 10-5 Air 80°C pH 7

10-6 Argon 120°C pH 5 0 2,000 4,0006,000 8,000 Time from beginning of irradiation phase (s)

Figure 2 Iodine fraction trapped on the molecular iodine selective filter, initially contained in the solution; effect of pH and temperature on iodine volatility for an initial iodide concentration of 10-5 mol/liter.

counters performed on-line γ spectrometry and, if necessary, develop new models to to measure activity trapped on the various improve assessment of the amount of filters and represent the kinetics of iodine volatile iodine involved in a severe accident specie formation. situation.

Five tests dedicated to iodide radiolytic Testing currently under way on EPICUR oxidation were conducted on IRSN's sets out to study the formation of organic EPICUR facility (Experimental Program on iodides from immersed painted surfaces Iodine Chemistry Under Radiation) at (simulating painted containment surfaces) Cadarache between May 2005 and July or from iodine-loaded paints placed in a 2007. Investigations covered the influence gaseous phase (simulating surfaces in the of oxygen presence, temperature, the ini- containment atmosphere where iodine may tial iodide concentration, and pH on be deposited). molecular iodine formation.

The results obtained confirmed that iodine volatility drops quickly when the solution pH changes from 7 to 5 (Figure 2). The solution temperature and the presence of oxygen have less effect on iodine volatility. The results obtained for initial iodide concentrations between 10-4 and 10-5 mol/liter indicate that this parameter has a relatively low impact on iodine volatility. These experiment results are being used to validate iodine behavior models in the French- German ASTEC code (Accident Source Term Evaluation Code), that serves in nuclear safety studies. The purpose is to identify the strong points and weak points of this code

130 2008 Scientific and Technical Report - IRSN 2.8 ASSESSING THE CONSEQUENCES OF MAJOR ACCIDENTS IN THE PRESENCE OF AIR: recent advances

Olivia COINDREAU Severe Accident Study and Simulation Laboratory Joëlle FLEUROT Major Accident Study and Simulation Laboratory Franck ARREGHINI Department for the Study and Modeling of Fuel in Accident Situations Christian DURIEZ Mechanics and Materials Experimentation Laboratory

Fuel rod cladding on pressurized water reactors constitutes the first containment barrier of radioactive substances present in the reactor core. Fuel cladding may be exposed to air under certain accident conditions, for example, extended uncovery of irradiated fuel assemblies during handling or storage in pools; or in a core meltdown situation, where it is exposed to air either after a reactor vessel failure, or in an accident that occurs while the reactor vessel is open. These accidents take place at "low" temperatures (600-1,000°C) and "high" temperatures (900-2,200°C), respectively.

How does air increase risk? formation of highly inflammable zirconium nitride, if oxygen in the air is completely consumed. The probability of a core meltdown accident occurring is extreme- The article below presents an accident transient and the associ- ly low, since it implies the loss of fuel cooling along with a partial ated phenomena. This description is followed by details on recent or total failure of safety systems. Nonetheless, when an accident research work carried out in international programs to further of this type does occur, exposure of zirconium alloy cladding to air knowledge of the corresponding phenomena. The last part is dedi- is one of the key physicochemical phenomena involved in accident cated to the description of the modeling enhancements realised in progression. the ASTEC computer code. When zirconium oxidizes through contact with the air, the reaction is highly exothermic (released energy representing 85% more heat than steam oxidation), and may lead to temperature runaway, with Flooding fuel assemblies: what are the destruction of the first containment barrier and acceleration of fuel risks and how can they be assessed? assembly degradation. Furthermore, the air oxidation reaction can lead to two other After having been irradiated in the reactor, fuel assemblies are consequences: immersed in water, first in the spent fuel pool located near the release of ruthenium tetraoxide, an extremely volatile compound reactor, then, once they have lost enough residual power, in a stor- [Mun et al., 2007]; age pool at the La Hague plant while waiting to be reprocessed.

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Underwater storage consists of placing each individual irradiated assembly in a compartment (integral with the pool bottom) that keeps the assembly in an upright position (Figure 1), even in the event of an earthquake.

Water drainage from a spent fuel pool or storage pool can expose the uncovered part of the stored assemblies to air in the building, depending on the accident scenario and the emergency measures taken by the operator to restore the correct water level in the pool. If the fuel assemblies are no longer cooled, the temperature in the uncovered part of the fuel rods could rise due to the decay heat from irradiation that the fuel continues to release.

The first risk associated with fuel assembly heating under accident conditions is failure of the first barrier that separates fission products from the environment. This can occur due to fuel rod cladding Figure 1 Irradiated fuel assembly storage rack in a US spent fuel reactor pit. failure, as the cladding loses its mechanical properties under the effect of internal pressure caused by the filler gas and fission gases. Studies on pool drainage accidents currently conducted at IRSN Two main heat exchange mechanisms have been identified that focus on the risks associated with air-oxidation of Zircaloy. These naturally limit fuel assembly heating during the accident. The first studies are conducted at three different levels: fuel rod cladding, is natural air convection produced in the storage building by cooled fuel assembly and storage, and finally, the spent fuel building and air located in the upper part of the building and warm air located storage pool. in the lower storage area. This convection mechanism can effectively cool the uncovered part of the storage compartments and The purpose of the studies is, first of all, to gain a better understand- the fuel assemblies they contain. The other exchange mechanism, ing of the physicochemical mechanisms that microscopically drive which occurs at higher temperatures, consists of heat radiation air-oxidation of Zircaloy, in which nitrogen plays a particularly impor- between the fuel assembly cladding and the compartments, on one tant role. This knowledge is based on interpretation of MOZART(1) hand, and between the compartments and the building and pool analytical test results, and is essential in building oxidation models structures, on the other. that take into consideration the various kinetic conditions of the oxidation reaction, which depends mostly on temperature and the The efficiency of cooling uncovered assemblies using natural air thickness of the existing oxide scale. By using experimentally vali- convection and heat radiation depends in particular on the storage dated oxidation models, it is possible to calculate the temperature configuration, the distribution of the assemblies and their residual thresholds at which temperature runaway may occur on uncovered heat, and the density of the storage compartment network. assemblies more accurately, to locate the zones where runaway occurs, and to predict runaway propagation conditions in the assemblies. In certain storage situations involving irradiated fuel assemblies, studies conducted by IRSN showed that the risk of a highly The study of air-oxidation conditions on uncovered fuel rods also exothermic reaction from Zircaloy cladding oxidation under air requires assessment of the thermal-mechanical behavior of fuel atmosphere is real, leading to fuel assembly temperature runaway, assemblies in the accident situation. The first calculations performed within time periods ranging from a few hours to several tens of using the ICARE-CATHARE code(2) [Fichot et al., 2006], for a fuel hours. The main risk associated with this thermal runaway assembly stored in its compartment or handled outside of the phenomenon involves first of all additional release of fission compartment, showed that fuel rod deformation induced by creep products or forms of uranium oxide volatilized through contact in the first temperature escalation phase could have a significant with the air. The other significant risk entailed by exothermic air-oxidation of fuel cladding is the embrittlement or even melting (1) MOZART: program to measure air-oxidation of zirconium as a function of tem- of certain metal compounds (Zircaloy, steel), jeopardizing perature. mechanical strength, cooling, and post-accident handling of the (2) ICARE/CATHARE: coupling between the mechanistic code for core meltdown ICARE 2, developed by IRSN, and the thermohydraulic code CATHARE 2, devel- fuel assemblies. oped by CEA, IRSN, EDF and Areva-NP.

132 2008 Scientific and Technical Report - IRSN Accidents in nuclear facilities 2.8

Parabolic rate constant (106 kg2/m4/s) Which research programs will improve 100,000 1,500 °C 1,300 °C 1,100 °C 1,000 °C 900 °C 800 °C knowledge and computer codes?

10,000 Scientific approach 1,000 A status report on the study of fuel rod cladding oxidation under 100 air atmosphere, conducted by IRSN in 2003 [Le Dantec, 2004]

10 highlighted, first, the considerable disparity between existing experimental results, and second, the non-representativity of the 1 cladding used in the tests, since it had never been exposed to reac- 0.1 tor conditions, and therefore bore no signs of corrosion (neither

0.01 oxidation, nor absorption of hydrogen from water). 5 678 9 10 Reciprocal temperature (104/K) Nonetheless, within the limits of existing tests, parabolic laws NUREG1 NUREG2 representing zirconium air-oxidation kinetics were established NUREGB AEKI (Figure 2), based on recommendations from the OPSA report(5) Figure 2 Correlations for constants in the parabolic oxidation laws [Shepherd et al., 2000] and used in computer codes to calculate recommended in the OPSA report. accident transients. During qualification of these codes through integral tests such as CODEX-AIT1 and QUENCH-10, tests simulat- influence on natural air circulation conditions in the compartments, ing an accident transient with air-oxidation on a bundle containing and therefore on the conditions that initiate and maintain the about twenty fuel rods clearly revealed the inadequacy of these zircaloy oxidation reaction. Calculations also showed that, given models [Duriez et al., 2008]. More specifically, it appeared that the the relatively low air circulation flow rates to be expected around models based on parabolic oxidation reaction kinetics did not take the stored assemblies (depending on the storage configuration and into consideration all the significant physical phenomena and were density), Zircaloy oxidation leads to the total consumption of oxygen not conservative. available in the air (starvation effect that limits temperature runaway). Below the runaway front in the assembly, the cladding In the context of the International Source Term Project (ISTP) coor- is immersed in an atmosphere containing primarily nitrogen, which dinated by IRSN, in cooperation with various national and interna- in the long term leads to the formation of zirconium nitride tional partners, the analytical test program MOZART was proposed, compounds that are mechanically very fragile and may be subject dedicated to studying oxidation of fuel cladding under air atmosphere to runaway once they are again in contact with oxygen. [Le Dantec, 2006].

Finally, studies and calculations have been planned for a macro- In parallel, research work to improve understanding and modeling scopic investigation representative of the entire storage rack in the of fuel cladding air-oxidation phenomena were conducted through pool, to assess how the accident progresses when runaway is initi- the SARNET(6) network of excellence [Micaelli et al., 2006], as part ated on an assembly where residual heat is particularly high and of the EU 6th Framework Program. In this context, network partners then propagated to other assemblies having lower residual heat. FZK, INR and IRSN conducted experimental programs [Duriez et These calculations are expected to call on the ASTEC code(3) [Van al., 2008], while ENEA, GRS, and other partners contributed to Dorsselaere et al., 2005], and the finite-element code CFX(4). enhancement of existing models [Coindreau et al., 2008]. The purpose of this work was to enrich the experimental database On a final note, IRSN has recently agreed to participate in an required to develop and qualify the models used in the ASTEC important experimental program directed by the US-NRC, carried integral code [Van Dorsselaere et al., 2005], the European reference out at the Sandia National Laboratory. Tests have been scheduled computer code used to study core meltdown accidents in water from 2009 to 2011 to study and measure the behavior of models reactors. representative of pressurized water reactor fuel assemblies in their storage environment when exposed to air at temperatures above (3) ASTEC: Accident Source Term Evaluation Code. 600°C. The results of these US tests will serve as the basis for (4) CFX: Computational Fluid Dynamix. validation of computer codes used at IRSN to study uncovery in (5) OPSA: Oxidation Phenomena in Severe Accident. spent fuel pool accidents. (6) SARNET: Severe Accident Research Network.

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Progress in understanding physical phenomena Mass gain (g/m2)

Within the MOZART experimental program, initiated in 2004 and Oven temperature (°C) Oxidation rate * 10 (μmZr02/h) pursued until the end of 2007, IRSN carried out small-scale oxida- 600 tion tests on zirconium alloy cladding. The test specimens, fuel 800 cladding sections from 1 to 2 cm long, were oxidized under air 500 600 atmosphere in isothermal conditions on a thermobalance. This tool Breakaway 400 is used for thermogravimetric analysis (TGA), i.e. it monitors mass 300 gain on the specimen in real time as it reacts to high temperature 400

in the ambient atmosphere. The oxidation rate was obtained by 200 taking the derivative on the mass gain signal time. Temperature in 200 100 the MOZART tests ranged from 600 to 1,200°C, covering the range in which temperature runaway and reaction runaway may be 0 0 -60 60 180 300 420 540 660 expected in an accident situation. After each test, metallographic Time (min) cross-sectional cuts were made on the specimens to interpret results. Air injected The various zirconium alloys used in French PWRs were studied:

Zircaloy-4 (tin-based alloy), M5 (a recent niobium-based alloy Figure 3 Example of an isothermal oxidation test on the thermobalance: oxidation at 800°C of a Zircaloy-4 specimen, developed by Areva), and Zirlo (tin- and nobium-based, marketed pursued until complete oxidation. by the American company Westinghouse). In the first phase, the cladding specimens were used in the new state, as received from the manufacturer. In the second phase the specimens were pre- bolic reaction constants were determined for each of the three oxidized at low temperature under steam before air-oxidation, to alloys investigated. They are presented in Figure 4, where they are simulate an initial corroded state representative of used assemblies. compared to different recommendations available in the technical In the final phase, a few tests were conducted to assess any influence literature to describe oxidation in the parabolic regime of zirconium of hydrogen dissolved in the metal, which is known to be present in alloys at high temperature, under steam and air atmospheres. In cladding at the end of its service life. the lower part of the tested temperature range, the experimental values obtained for Zircaloy-4 were noticeably lower than those of This small-scale test program made it possible to compile a detailed the other two alloys, corresponding to slower diffusion of oxygen in database on air-oxidation kinetics, while converging towards a the zirconia layer. The change in gradient observed at 1,000°C cor- better understanding of the phenomenology behind zirconium alloy responded to the monoclinic/quadratic phase boundary in the degradation under air atmosphere, which pointed to the funda- zirconia. On an overall basis, the data obtained were closer to the mental role played by nitrogen [Duriez et al., 2008; Duriez et al., "steam" kinetic correlations than the very conservative recommenda- 2009]. tions given in the OPSA report [Shepherd et al., 2000], commonly A typical result from thermogravimetric analysis on the new clad- used in major accident computer codes to describe oxidation under ding specimen is shown in Figure 3. After a temperature excursion air atmosphere. under argon, mass gain started when air sweeping began in the oven. This test ran until the metal was completely oxidized, indi- At a thickness that depends on temperature and type of alloy, the cated by a steady mass gain signal. Two distinct kinetic regimes protective zirconia layer cracked due to the effect of high compres- were identified on the derived signal. sive stress at its center, which gave way to an acceleration of oxidation. Cracking was associated with a zirconia transformation from The first, referred to as the pre-transition regime, was characterized the quadratic phase to the monoclinic phase when stress was by an intense peak due to oxidation of the bare metal, followed by relieved, inducing microcracking of the oxide scale, contributing to a decrease in the oxidation rate caused by a protective layer of attenuation of its protective role. This complex phenomenon of oxide scale, which gradually grew thicker. In this case the solid-phase cracking and phase transformation is referred to as "breakaway". diffusion of oxygen through the zirconia is the layer up to the metal/ The thermogravimetric technique determined the precise moment oxyde interface limiting step of the reaction. In an initial approach, when kinetics changed, along with the corresponding thickness, the regime could be described by a parabolic law such as Δm/S = k calculated based on the minimum mass gain in the oxidation rate t1/2, where k is a constant called the parabolic reaction constant curve. The thickness measured at the kinetic changeover point for and Δm/S is the mass gain over the metal unit of area. The para- the three investigated alloys ranged from a few micrometers at low

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temperature to several tens of micrometers at 1,000°C (Figure 5).

Above 1,000°C, breakaway was no longer observed, the protective 200°C 100°C 000°C 1.101 1, 1, 1, 900°C 800°C 700°C 600°C oxide layer growing to significant values, and the kinetic regime NUREG-2 remaining parabolic. 1.10

Once breakaway occurred, the post-transition regime was practically linear at low temperatures (up to 700°C). After a rapid rise, 1.10-1 the oxidation rate stabilized at a constant value. But starting from NUREG-1

800°C, the oxidation rate did not stop rising after the transition. The 1.10-2 metallography images show the formation of zirconium nitride (ZrN) Leistikow-Schanz particles near the metal/oxide interface, dispersed in a porous 1.10-3 structural oxide (Figure 6). ZrN was initially formed at the bottom 6781910 112 of the cracks in the dense oxide layer, where the oxygen was consumed and nitrogen had accumulated. This revealed local oxygen Zry-4 M5 Zirlo starvation at the microscopic scale. The oxidation front continued to progress towards the inside of the material and the ZrN particles Figure 4 Parabolic reaction constants determined based on isothermal kinetics in the pre-transition range. Comparisons were converted into oxide. The high increase in volume associated with NUREG recommendations for air-oxidation and with with this conversion generated oxide cracking, making the oxide the Leistikow-Schanz recommendation for steam-oxidation. scale porous and eliminating its protective role. Once it was initiated, this growth of porous oxide under the effect of nitriding was self-fed, since the nitrogen released through nitride oxidation remained trapped in the material and was available to resume Oxide thickness at breakaway (μm) formation of nitride particles from the metal. The significant metal 100 creep generated by this nitriding/oxidation mechanism, associated with its self-feeding quality, explained the continuous growth 80 of the cladding degradation rate observed after breakaway. 60 The presence of an initial layer of corrosion favored the transition to this nitro-oxidation regime, which resulted in a decrease in the 40 duration (or even the absence) of the parabolic regime for specimens pre-oxidized at low temperature. Nonetheless, while the layer of 20 corrosion was thick (approx. 60 to 80 µm), it played a protective role with regards to the oxygen supply, making degradation of the 0 600 700 800 900 1,000 specimen in the final phase slower than for a new specimen. Temperature (°C)

Zry-4 in M5 in Zirlo in received state received state received state

Figure 5 Thickness of the protective oxide layer at breakaway for the three investigated alloys.

Zirconium Porous oxide nitride Dense protective oxide

Cracked dense oxide

Figure 6 Metallography image of a Zircaloy-4 specimen oxidated under air atmosphere at 850°C.

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Moreover, in an accident situation, the nitrogen-rich atmospheres the ICARE code could reliably predict the temperature trend for the that may be found locally would also encourage this transition to pre-oxidation phase (under a mixture of steam and argon), but did a nitrided regime. not correctly reproduce the air-oxidation phase, underestimating temperature runaway, in particular. The zirconium air-oxidation reaction therefore appears as a complex phenomena, where nitrogen plays a predominant role by consider- It therefore appeared necessary to improve this model of oxidation ably accelerating degradation. Given the highly exothermic nature under air atmosphere. A study on experimental results from the of the zirconium reactions with oxygen and nitrogen, it appears easy MOZART program and tests carried out in the SARNET program to understand why fuel assembly exposure to the air can lead to led to a new model that simulates the phenomena occurring during much greater runaway than exposure to steam. It is therefore the fuel rod cladding air-oxidation reaction, to differentiate oxidation essential to take these phenomena into consideration in simulation kinetics before and after breakaway, and also to evaluate the instant codes designed for severe accidents (such as fuel assembly uncov- when this transition occurs [Coindreau et al., 2008]. ery in spent fuel storage pools, reactor core meltdown accidents). The new model thus assesses an oxidation regime before break- Modeling enhancements away, through parabolic oxidation kinetics that describe total Initially, the ICARE code, a module of the ASTEC integral code simu- mass gain as a function of square root of time. The model can lating the behavior of fuel assemblies during an accident transient, use different correlations to calculate these oxidation kinetics, partially took into account air-oxidation phenomena by using the taken from NUREG1-2(7) [Powers et al., 1994], NUREGB and AEKI parabolic oxidation reaction kinetics taken from OPSA recom- [Shepherd et al., 2000], as well as a correlation based on MOZART mendations [Shepherd et al., 2000], to assess growth of the zirco- results. nia (zirconium oxide) layer. Results of qualification runs of this model in the QUENCH-10 test [Mladin et al., 2006] showed that (7) NUREG: Nuclear Reactor Regulation.

600°C 700°C Mass gain (g/m2) Mass gain (g/m2) 700 700

600 600

500 500

400 400

300 300

200 200

100 100

0 0 0 10,000 20,000 30,000 0 1,000 2,000 3,0004,000 Time (min) Time (min)

800°C 1,000°C Mass gain (g/m2) Mass gain (g/m2) 700 700

600 600 New model 500 500

400 400 Experiments 300 300 Old model 200 200

100 100

0 0 0 100 200 300 400 500 600 700 0 20 40 60 80 100 Time (min) Time (min)

Figure 7 Comparison between MOZART experimental results (brown and pink curves) and the mass gain curves calculated using ICARE/CATHARE: standard oxidation model (parabolic law) in blue, new model taking into account the kinetic transition (breakaway) shown as a solid black line.

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In an evaluation on the moment of breakaway, MOZART test results Further progress still lies ahead. Regarding the transition from showed that a transition occurred between the parabolic oxidation one oxidation regime to another, it appears that additional tests phase and the accelerated oxidation phase for a critical mass gain are necessary to obtain a more accurate characterization of this value that depends to a great extent on temperature. It also transition. Current models process the different phenomena appeared that mass gain at the instant of transition could be that influence the oxidation rate in the post-transition phase using correlated to temperature by a hyperbolic law, based on the assump- accelerated kinetics. Analysis of experimental results has shown, tion that the kinetic transition is related to a phase change in the however, that the zirconium alloy fuel rod cladding oxidation zirconia layer from a quadratic to a monoclinic phase. kinetics seems to be strongly related to the mechanical behav- After breakaway, the new model described accelerated oxidation ior of the zirconia layer. It is therefore necessary to develop a kinetics for total mass gain using an empirical law based on the new detailed model simulating the various important phenomena experimental results. that impact the state of stress in this zirconia layer, such as, for example, an increase in layer volume, the presence of nitride at where the metal/oxide interface, and oxidation of the ZrN formation. To appreciate the contribution of this new model, the isothermal Furthermore, current models must be extended to cover new MOZART tests were calculated using the old model and the new cladding materials such as Zirlo and M5. model. A comparison between experimental results and calculated results (Figure 7) showed that the time required for complete cladding oxidation estimated using the old model is largely over- estimated compared to the new model, which predicts values very close to the experiment observations, and is therefore more conservative in terms of the oxidation rate and the temperature excursion rate.

Conclusions and outlook

Research programs on fuel assembly cladding oxidation under air atmosphere conducted in the last few years have contributed knowledge on the physical phenomena involved. More specifically, new experimental data obtained in the ISTP and SARNET international programs have revealed the following findings: the commonly used parabolic oxidation kinetics taken from OPSA recommendations only describe a limited portion of the cladding degradation transient; the transition from a parabolic oxidation regime to an accelerated oxidation regime must be taken into consideration.

To acknowledge all of these observations, the models used in accident transient simulation codes have been enhanced and the first applications show a better predictive capability when compared with experimental results. These improvements constitute an important advance in simulation computer codes, which should now be capable of producing more realistic assessments of the oxidation kinetics and temperature runaway phenomena induced by accident situations where fuel rod cladding is exposed to air.

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References O. Coindreau et al. (2008). Modelling of accelerated cladding degradation in air for severe accident codes, ERMSAR, Nesseber, Bulgaria, 23-25 septembre 2008. C. Duriez, T. Dupont, B. Schmets, F. Enoch (2008). Zircaloy-4 and M5® high temperature oxidation and nitriding in air, J. Nucl. Mat. 308, p. 30-45. C. Duriez, M. Steinbrück, D. Ohai, T. Meleg, J. Birchley, T. Haste (2009). Separate-effects tests on zirconium cladding degradation in air ingress situations, Nucl. Eng. Design 239, p. 244-253. F. Fichot et al. (2006). Multidimensional approaches in severe accident modelling and analyses, NET. G. Le Dantec (2004). Étude bibliographique sur l’oxydation des gaines de combustible en zircaloy sous air, Note technique DPAM/SEMCA 2004-25. G. Le Dantec (2006). Separate effect tests on oxidation of fuel cladding by air. Definition of the needs, Source Term Program, ISTP Report N° 14. S. Leistikow, G. Schanz (1987). Oxidation kinetics and related phenomena of Zircaloy-4 fuel cladding exposed to high temperature steam and hydrogen-steam mixtures under PWR accident conditions, Nucl. Eng. Des. 103, p. 65-84. J.C. Micaelli et al. (2006). SARNET. A European Cooperative Effort on LWR Severe Accident Research, Revue Générale Nucléaire, n° 1, January-February 2006. M. Mladin, G. Guillard, F. Fichot (2006). Post-Test Analysis of QUENCH-10 Experiment with ICARE/CATHARE, In 12th Int. QUENCH Workshop, FZ Karlshrue, October 2006, ISBN 976-3-923704-57-6. C. Mun et al. (2007). Étude de la chimie du Ruthénium dans l’enceinte de confinement en cas d’accident grave, IRSN, Rapport scientifique et technique. D.A. Powers, L.N. Kmetyk, R.C. Schmidt (1994). A Review of Technical Issues of Air Ingression during Severe Reactor Accidents, NUREG/CR-6218, September 1994. I. Shepherd et al. (2000). Oxidation Phenomena in severe Accidents (OPSA), INV-OPSA (99)-P008, EUR 19528 EN. J.P. Van Dorsselaere et al. (2005). ASTEC and SARNET, Integrating severe accident research in Europe, ICAPP’05 (15-49 mai), Séoul (Corée). http://sarnet.grs.de/default.aspx.

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Key events and dates

Dissertations OTHER defended major events

February 14, 2008 February 2008 Elie Chalopin submitted a thesis entitled COPERNIC Platform "Characterization of the radiative properties of February 5, 2008: Label awarded by the PACA a porous medium using the RDFI method – region's Territorial Risk and Vulnerability Application to a degraded reactor core" in Management Cluster for the Copernic joint Paris. platform on fire knowledge, led by IRSN. This platform allows companies to call on IRSN's September 24, 2008 competence and test facilities in areas involv- Florence Gupta submitted a thesis on the ing fire risk. "Study of cesium behavior as a fission product in uranium dioxide using the ab initio method" April 2008 in Cadarache in southern France. Patent application submitted for a device to measure surface displacement and its applica- October 29, 2008 tion to measuring the regression of the surface Sébastien DESTERCKE submitted a thesis of a combustible material that is optically trans- in the field of information processing in the parent in a fire. presence of uncertainty, entitled "Methods of synthesis of imprecise probabilistic informa- May 2008 tion" at the Université Paul Sabatier in CHIP facility commissioned Toulouse. May 14, 2008 saw the commissioning of CHIP, the new facility designed to quantify and characterize radioactive iodine release in the event of core meltdown in a nuclear reactor. The experimental CHIP program (to study iodine chemistry in the reactor coolant system), with its new test facility located at Cadarache in southern France, aims to increase assessment capabilities at the Institute in the field of emergency prevention and emergency response management applied to the environmental release of radioactive iodine.

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