Energy Conversion and Management 64 (2012) 522–529
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Energy Conversion and Management
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Neutronic characterization and decay heat calculations in the in-vessel fuel storage facilities for MYRRHA/FASTEF ⇑ S. Di Maria a, , M. Ottolini b, E. Malambu Mbala c, M. Sarotto d, D. Castelliti c a Instituto Tecnológico e Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, EN 10, 2686-953 Sacavém, Portugal b Ansaldo Nucleare S.p.A., c.so Perrone 25, 16161 Genova, Italy c SCK–CEN, Boeretang 200, Mol B-2400, Belgium d ENEA, via Martiri di Monte Sole 4, 40129 Bologna, Italy article info abstract
Article history: The main objective of the Central Design Team (CDT) project is to establish an engineering design of a Received 19 October 2011 Fast Spectrum Transmutation Experimental Facility (FASTEF) that is the pilot plant of an experimental- Received in revised form 3 May 2012 scale of both an Accelerator Driven System (ADS) and a Lead Fast Reactor (LFR), based on the MYRRHA Accepted 3 May 2012 reactor concept, planned to be built during the next decade. The MYRRHA reactor concept is devoted Available online 16 August 2012 to be a multi-purpose irradiation facility aimed at demonstrating the efficient transmutation of long-lived and high radiotoxicity minor actinides, fission products and the associated technology. An important Keywords: issue regarding the reactor design of the MYRRHA/FASTEF experiment is the In-Vessel Fuel Storage Facil- Fast reactor ities (IVFSFs), both for fresh and spent fuel, as it might have an impact on the criticality of the overall sys- Neutronic design Decay heat tem that must be quantified. In this work, the neutronic analysis of the in-vessel fuel storage facility and its coupling with the critical core was performed, using the state of the art Monte Carlo program MCNPX 2.6.0 and ORIGEN 2.2 computer code system for calculating the buildup and decay heat of spent fuel. Sev-
eral parameters were analyzed, like the criticality behavior (namely the Keff), the neutron fluxes and their variations, the fission power production and the radiation damage (the displacements per atom). Finally, also the heat power generated by the fission products decay in the spent fuel was assessed. Ó 2012 Elsevier Ltd. All rights reserved.
1. Introduction reactor, and at the present status of the design the core consists of MOX fuel pins with an active length of 60 cm and 33% of pluto- The aim of the CDT project [1,2] is to further develop the engi- nium content [1]. The choice of the MOX fuel was made taking into neering design of a first-step experimental device based on the account the large experience in the MOX production in Europe and resulting MYRRHA/XT-ADS facility of the FP6 EUROTRANS project its neutronic properties for a fast spectrum. that may serve both as a test-bed for transmutation and as a fast An important issue regarding the MYRRHA/FASTEF nuclear spectrum irradiation facility, operating as a subcritical [3] in Accel- reactor design is the In-Vessel Fuel Storage Facility (IVFSF), both erator Driven System (ADS) mode [4–6], and/or as a critical reactor. for fresh and spent fuel: in the current design, in order to avoid The CDT project defines also the new R&D activities needed to aid excessive delays between two operation cycles, it was decided to the detailed design and the construction of such facility that is place the IVFSF at the reactor periphery. This is a peculiar feature planned to be built during the next decade under the supervision of MYRRHA/FASTEF, since most of the present working reactors ex- of the SCK CEN Belgian Nuclear Research Centre and in collabora- ploit fuel storages not inserted in the reactor vessel. For this reason, tion with different national and international partners. MYRRHA/ it is crucial to assess the neutronic interaction between the central FASTEF, in the subcritical operational mode, has a proton accelera- core and the Fuel Assemblies (FAs) in the storage zones during the tor with energy of 600 MeV and a maximal current intensity of reactor operations, both in terms of criticality estimation and in 4 mA, coupled to a liquid lead bismuth eutectic (LBE) spallation terms of material damage to the overall structure. source. The spallation target is located at the center of the subcrit- As for the fresh fuel, it is important to know the neutron cou- ical core (see Fig. 1). The LBE coolant adopted in all the primary pling between the critical core and the storage vessels, the power system allows a fast neutron spectrum. The reactor is a pool-type generated in the storage vessels due to the fission induced by the neutron flux coming from the core and the structural damage of the vessel materials. Since a large amount of heat continue to be ⇑ Corresponding author. developed even after a reactor has been shutdown, it is essential E-mail address: [email protected] (S. Di Maria). to know the exact heat power released (i.e. by fission product
0196-8904/$ - see front matter Ó 2012 Elsevier Ltd. All rights reserved. http://dx.doi.org/10.1016/j.enconman.2012.05.001 S. Di Maria et al. / Energy Conversion and Management 64 (2012) 522–529 523
Fig. 1. Section view of the MYRRHA/FASTEF design operating both in critical and subcritical mode. decay) in this phase and remove it in a safe way, otherwise the fuel tions (IPSs) locations foreseen for irradiation applications [8].As elements may suffer damage (e.g. melting). shown in Fig. 2, the core design envisages a total of 151 positions In this paper the state of the art of the IVFSF for the MYRRHA/ and among them 37 penetrations (in the legend indicated as FASTEF design is described. Several parameters will be analyzed, Multi-Functional Channels) are planned to host indifferently IPS, as the criticality behavior (Keff), the fission power production and Control Rod (CR), Scram Rod (SR), FA and Dummy Assemblies the radiation damage, namely the displacement per atom (dpa). Fi- (DAs) [8]. nally, also the decay heat power of spent fuel generated for differ- The central penetration is used for the beam tube/target in sub- ent operational reactor periods will be assessed. critical mode and can be used for an IPS (or CR/SR) in critical mode. The calculations reported here were done by considering the It was decided to include four IVFSF, each one containing 76 reactor operating in critical mode. assembly positions that will be able to store a maximum of two full core loads. There are two IVFSF for each In-Vessel Fuel Handling 2. Materials and methods Machines (IVFHMs). The FA are inserted in the pipes of the IVFSF and remain there until their residual heat has sufficiently decayed. The core model used for our calculations is the LBE-filled box The positions of the IVFSF are also locations of fresh FA for refuel- dummy and the layout core design is shown in Fig. 2 [7]. The core ing and DA. The two IVFHM are positioned at opposite sides of the is composed of 68 fuel assemblies with a total power of 100 MW in core, each capable of accessing half of the core and half of the stor- the critical operational mode (in subcritical mode the operational age positions, thus allowing to minimize the diameter of the reac- power will be in the range 65–100 MW). The design goal of the tor vessel (see Fig. 3). The IVFSF are placed at the same level of the core is to achieve the maximum of the flexibility in the In-Pile Sec- core and far from it in order to avoid any neutron coupling [9].
Fig. 2. Left: Preliminary design of the 100 MW (68 FA) critical core; in the legend, the different positions for fuel and dummy assemblies, B4C rods and IPS are indicated. In particular the last two parenthetical positions indicate alternative positions (see text for details). Right: MCNPX critical core model used for calculations. 524 S. Di Maria et al. / Energy Conversion and Management 64 (2012) 522–529
Fig. 3. Overall top view of the MYRRHA/FASTEF Diaphragm design with all the parts included in the reactor vessel (see text for details).
The core radius is 72 cm and the distance between the core cen- In our case Efiss is about 207 MeV/fission (taking into account ter and IVFSF center is 258 cm. For neutron calculations the both prompt and delayed energy release). Since the power of the MCNPX 2.6.0 Monte Carlo code [10] was used together with the core is known, we can derive the Normalization Factor (NF): JEFF 3.1 nuclear data libraries [11].InFig. 4 the MCNPX MYR- PðWÞ RHA/FASTEF design is shown: in the MCNPX model, at this stage, NF ¼ ð1Þ 13 J MeV Nfiss only the four IVFSF, critical core and diaphragm (part of the reactor 1:6022 10 MeV 207 fiss s that divides the hot and cold LBE zones), were taken into account to perform criticality calculations. where Nfiss is the number of fissions per second in the core. Know- It is not essential to include other components of the system, ing the partial Nfiss of the IVFSF, it is straightforward to obtain the such as the Primary Pump or the Primary Heat Exchanger, since power released in the storage vessel. they are situated far from the active core and they would not influ- Many radiation-induced effects may cause material damage, as ence the criticality. for instance direct heating and production of lattice defects. In fact, The barrel surrounding the core, IVFSF and diaphragm, are com- it has been shown that there is an empirical correlation between posed of AISI 316L Stainless Steel of 2 cm, 1.5 cm and 4.5 cm thick- the number of dpa and various properties of materials [12]. In this ness respectively. Different parameters were estimated by MCNPX, work, the estimation of the neutron flux coming from the critical core was undertaken, in order to study a possible structural dam- namely the Keff value, the neutron fluxes and reactor power. The number of fission, in terms of neutrons lost by fission per age in the vessels containing the stored fuel assemblies. To this purpose the following formula was used [13]: source neutron, is contained in the summary balance table of Z MCNPX, and it is given cell-by-cell. Besides it is possible to esti- dpa hrEi mate the total average recoverable energy by fission, once the iso- ¼ g /ðEÞdE ð2Þ s E 2Ed topic composition of the fuel is provided.
Fig. 4. MYRRHA/FASTEF MCNPX model used for calculations (core & four IVFSF). The particular of the IVFSF with the 76 positions is also shown. S. Di Maria et al. / Energy Conversion and Management 64 (2012) 522–529 525
where Ed is the required energy to displace an atom from its lattice Table 1 position (in the range 25–40 eV for SS 316L), g = 0.8 is the displace- Keff study according to three different shapes of the storage vessel. ment efficiency, hrEi is the dpa energy cross-section (in this work Keff (real) Keff (semi cylindrical) Keff (elliptical) we considered SS 316L as a compound of Ni58, Ni60, Cr52, Fe54 0.93421 ± 0.00031 0.93845 ± 0.00051 0.93807 ± 0.00061 and Fe56) computed by the data processing modules of NJOY [14,15] and U(E) is the incident neutron flux per energy bin, calcu- lated by the F4 cell flux tally with MCNPX and the NF in Eq. (1). Another important issue in the design of shielding for handling Table 2 Keff values for three core configurations (with and without FA in each of the four high level waste is the production rate of neutrons from spontane- IVFSF). ous fission (SF) and (a,n) reactions. Neutron production from (a,n) reactions takes place as a result of the action of energetic a parti- Core only at BOL Core + 68 FA in one of the Core + 17 FA in each of four IVFSF four IVFSF cles from a-decay of radionuclides such as Pu238, Am241, Cm242 and Cm244 with elements such as oxygen and fluorine. The a par- 1.04342 ± 0.00051 1.0453 ± 0.00059 1.04384 ± 0.00061 ticles lose energy very rapidly when traveling through matter and in the case of Uranium and Plutonium oxides the range is roughly 0.006 cm and 0.007 cm, respectively [16]. In many cases this short Table 3 range means that the a particles never reach nearby materials in (a,n) Neutron source intensity. which (a,n) reactions can take place. However in MOX-like fuel U235 37.8 n/s the contribute of (a,n) source could be non-negligible, since in this U238 60.99 n/s case the oxygen is intimately mixed with the a-emitting nuclear Pu238 1.14 108 n/s 6 material. Pu239 7.89 10 n/s Pu240 1.41 102 n/s In order to take into account these two neutron sources, as well Pu241 1.38 107 n/s as the decay heat power in the IVFSF, the ORIGEN 2.2 [17] program Pu242 5.6 104 n/s was used. In this type of program the neutron cross-section re- quired for the solution of the Bateman equations [18] are provided externally as libraries. Once the nuclide concentrations as a func- tion of the time are known, the decay heat after the reactor shut- Table 4 down can be obtained by multiplying the nuclide activities (a Spontaneous fission neutron source intensity. total of 1700 nuclides are present in the ORIGEN database) by their U235 1.59 n/s respective recoverable energy per decay values and summing over U238 9.99 103 n/s 7 the entire nuclide vector. Pu238 2.19 10 n/s Pu239 4.51 103 n/s Pu240 1.00 108 n/s 3 3. Results and discussions Pu241 1.09 10 n/s Pu242 4.81 107 n/s 3.1. Criticality assessment
To be conservative, the FA pitch in the vessel storage was de- Table 5 Power contribution in the IVFSF coming from three different neutron sources. signed to keep its Keff value lower than 0.95 when 76 fresh FA are loaded. The FA pitch in the vessel storage is composed by: Induced fission power from neutrons Spontaneous (a,n) source the FA with a 2 mm thick T91 wrapper, 6.475 mm of LBE (to avoid leaving the core fission source contact under swelling), 3 mm of AISI 316L (duct) and an external 1.5 MW 2.5 10 2 W 2.0 10 2 W LBE layer. To satisfy the aimed subcriticality level, a parametric study was performed by varying the thickness of the external LBE around the duct (see Fig. 5). By MCNPX simulations it was ob- tained that a fuel pitch of 156.5 mm maintains the storage vessel to Furthermore, to study the influence of the casing shape of the the aimed subcriticality (K = 0.93421). eff IVFSF, different shapes were considered (semi cylindrical and ellip- tical casing). The results pointed out that only a maximum varia- tion in the multiplication factor value of about 400 pcm with respect to the real shape considered can be addressed (see Table 1). For the criticality calculations we considered a conservative value of T = 300 K for the fuel and structural materials, as well as for neu- tron cross-sections in the JEFF 3.1 libraries. To study the neutronic coupling between the critical core and the IVFSF, different configurations were investigated. In particular, two possible scenarios were analyzed for the MYRRHA/FASTEF de- sign: in the first we considered the most conservative configura- tion with 68 FA in one storage vessel and 68 FA in the critical core, while in the second 17 FA are placed in each vessel and 68 FA in the critical core. This last configuration simulates the situa- tion where another core fuel load is distributed in the four storage vessels. Since with MCNPX it is not possible to separate the critical- ity contribution due to one neutron source only, we studied the multiplication factor variation for the whole system (68 FA in the core plus 68 FA in the IVFSF), first calculating the multiplication Fig. 5. Parametric study that shows the Keff dependence by the IVFSF pitch (in the x- axis of the graph the half-pitch is shown). factor for the critical core and then considering the system core 526 S. Di Maria et al. / Energy Conversion and Management 64 (2012) 522–529
Fig. 6. MCNPX visualization of the storage vessel showing the places where the dpa values were calculated. For comparison also the dpa value in the core barrel was reported.
Table 2 shows different values of Keff relative to the different configurations: since there is no appreciable variation, it is possible to conclude that the core and the IVFSF are not coupled. This means that the presence of fresh fuel assemblies in the vessel storages does not influence the multiplication factor of the critical core.
3.2. Power calculations
The contributions to the IVFSF heating have been calculated by taking into account both the neutron steaming from the reactor core and the local SF and (a,n) neutron sources, whose intensities were determined by the inventory of actinides and were obtained from ORIGEN2.2. The sources are shown in detail for each nuclide in Tables 3 and 4. The heat deposition in the IVFSF was evaluated using the F6 tally (that takes into account the energy deposition averaged over a cell) with MCNPX. As displayed in Table 5 the main contribution to the power in the storage vessels is due to the fission power generated by the Fig. 7. Axial distribution of dpa calculation in the storage casing. The point at z =0 neutron flux coming from the core. corresponds to the core middle plane and refers to the dpa value of 0.08 dpa/year in Fig. 6. 3.3. Displacement per atom calculations plus IVFSF as a whole. Thus, the eventual variation of the multipli- cation factor between the two configurations (core only and core The effect of the neutron flux on the SS316 L storage casing in plus IVFSF) should entail the neutronic coupling between core terms of dpa was also studied. Fig. 6 shows the dpa values of the and IVFSF. most exposed places to the neutron flux coming from the critical
Fig. 8. Neutron spectra calculated in the storage vessel (left) and in the core barrel (right). S. Di Maria et al. / Energy Conversion and Management 64 (2012) 522–529 527
Fig. 9. JEFF 3.1 Fe56 damage-energy cross section [19].
Fig. 10. Decay heat power calculated for five different reactor operational periods Fig. 11. Decay heat power calculations in the storage vessel for 68 FA performed and for 68 FA in the core. Time after shutdown is in logarithmic scale. considering a fuel cycle composed by five sub-cycles. In each sub-cycle about 20% of the fuel is changed every 90 days. Time after shutdown is in logarithmic scale.
Table 6 Decay heat values for two different reactors operational periods (68 FA). order of magnitude less than the one hitting the core barrel 2 Irradiation time 90 days Irradiation time 450 days (6.75E14 n/cm /s) and it is composed by neutrons with energies (MW) (MW) less than 1 MeV, whereas about 2% of the neutron spectrum in the core barrel has an energy greater than 1 MeV. In Fig. 9 the 30 s after shutdown 5.02 5.17 48 h after shutdown 0.44 0.54 Fe56 damage cross section is shown, because it is the nuclide with 300 days after 0.0175 0.041 the highest weight fraction in the SS316L. shutdown 3.4. Decay heat calculations core (z = 0 axial position corresponding to the core middle plane), Even considering long periods after the reactor shutdown while Fig. 7 displays the dpa axial distribution (from 60 to (when the fuel assemblies are transferred to the IVFSF), quite large +60 cm axial position). This kind of study was performed only for amount of heat continues to be developed in the fuel because of the part of the vessel with higher dpa value (0.08 dpa/year, the presence of fission products (decay heat power) [20]. Fig. 6). For comparison, also the damage in the core barrel was cal- The knowledge of the decay heat curve is important to design culated. The dpa in the core barrel is 2.8 dpa/year, about two orders the cooling system of the IVFSF. Anyway the influence of the decay of magnitude greater than the dpa value evaluated in the point 1 of heat on the cooling system and in general on the thermo-hydraulic the storage vessel (see Fig. 6). The difference between the two dpa design of the IVFSF is beyond the scope of this paper. values is mainly due to the different neutron spectrum hitting the In this work we considered the burn up of the 68 FA for five dif- core barrel and the storage vessel. In fact, as shown in Fig. 8, the ferent irradiation periods, calculating the decay heat power in the neutron flux hitting the storage vessel (7.85E13 n/cm2/s) is one core with ORIGEN 2.2, as shown in Fig. 10. 528 S. Di Maria et al. / Energy Conversion and Management 64 (2012) 522–529
Fig. 12. Left: Comparison between ORIGEN 2.2 and empirical formula decay heat power curves. Right: The deviation in percentage of the empirical formula values from the ORIGEN ones is shown. Time after shutdown is in logarithmic-scale.
It is possible to see how the power increases with the irradia- Moreover the fission yield distribution for any actinide depends tion time. In particular considering the first 60 s and 48 h after on the energy of the fission-inducing neutrons, therefore the con- reactor shutdown, the decay heat power for the 450 days of fuel centration and variety of fission products can be different for reac- burn up case is higher of about 3–4% and 18% respectively with re- tor system operated at the same nominal power, but with different spect to the 90 one, while for a time period of 300 days after shut- fuel composition and neutron energy spectra. down the decay heat power for the 450 days of fuel burn up is Another factor of discrepancy between the two curves can be higher of about 50% with respect to the 90 one (see Table 6). due to the different fission energy considered. In fact, while the Since the MYRRHA/FASTEF design was conceived to be a flexible empirical formula uses a constant recoverable energy per fission irradiation facility, for example for radioisotope production or for of 200 MeV, in ORIGEN it is assumed to be a function of the fission- neutron silicon irradiation purposes, the capability to get a reactor ing nuclide. In particular, in the case of constant recoverable en- core ruled to a frequent fresh/spent fuel replacement and shuffling ergy per fission, the flux required to sustain a given amount of strategy should be a peculiarity. For this reason we calculated the power could be not accurate, and thus the amount of the burned decay heat power after a potential complete fuel cycle. In a MYR- material could be not precise enough. However, considering that RHA/FASTEF fuel cycle operating in critical mode, about 20% of the empirical formula is accurate within ±50% for the first period the spent fuel will be replaced by fresh fuel in each sub-cycle, so [21,22], it is possible to see that deviations between the two curves a complete fuel cycle will be reached after 450 days. Generally, for the first 120 days after shutdown are small, whereas there is a after 450 days, the 68 spent FA placed in one IVFSF will have differ- divergence for longer periods. ent burn up and decay heat power values. Fig. 11 shows the decay These differences could also be explained also for a major con- heat power after a complete fuel cycle where the 68 spent FA in tribution coming from the actinides with a long decay period in one storage vessel are composed by: fast reactors with respect to the LWR ones. Anyway the compari- son of the two methods above described is only to show the impor- 13 FA with a burnup of 90 days and a decay heat of 360 days; tance of using the decay heat data derived from an appropriate 13 FA with a burnup of 180 days and a decay heat of 270 days; reactor model [18]. 13 FA with a burnup of 270 days and decay heat of 180 days; 13 FA with a burnup of 360 days and decay heat of 90 days; 16 FA with a burnup of 450 days. 4. Conclusions
The decay heat curve showed in Fig. 11 corresponds to the spent The main objective of this work was to perform systematic fuel composition above described (since it depends also on the studies and accurate estimations of important parameters such number of FA burned in each sub-cycle). The decay heat power as neutron coupling, material damage and decay heat power in was calculated without considering any particular fuel shuffling the in-vessel fuel storage facilities for the MYRRHA/FASTEF reactor strategy, and in general under the assumption of constant neutron design in the framework of the CDT project. flux irradiation in each sub-cycle. Concerning the neutron coupling between critical core and FA Since no standard procedure is actually available for fast reac- in the storage vessels, the fuel pitch of the IVFSF was estimated tors [18] and only for sake of comparison, we compared two decay through MCNPX Monte Carlo simulations in order to reach a Keff heat curves, one obtained with an empirical formula (equation value lower than 0.95. With the evaluated fuel pitch in the IVFSF, 2.57 in [21]) used for LWR reactors, the other calculated consider- even when the critical core is in operation and additional 68 fresh ing a fast reactor model with ORIGEN 2.2. In Fig. 12, a comparison FA assemblies are located in the storage vessel, there is no signifi- between the two decay heat distributions obtained for 90 days cative variation in the critical Keff value of the whole system. irradiation of 68 FA is shown, together with the difference in per- Regarding the heat power in the IVFSF, we assessed the effect of centage between the one obtained with ORIGEN and the other by three different neutron source contributions: the neutron flux empirical calculations. The differences here reported can be ex- coming from the critical core, the spontaneous fission and (a,n) lo- plained considering that the empirical formula takes into account cal sources. The results show that the main contribution to the heat only four fissionable nuclides (U235, U238, Pu239, Pu241) induced power in the IVFSF comes from the core neutron flux (about by thermal neutrons, while in our case the contribution coming 1.5 MW). The power heat generated by the fission products decay from other actinides is not negligible. in the spent fuel was calculated through the ORIGEN 2.2 depletion S. Di Maria et al. / Energy Conversion and Management 64 (2012) 522–529 529 program. Since MYRRHA/FASTEF will be a flexible irradiation facil- [2] Collaborative Project ‘‘Central Design Team (CDT) for a Fast Spectrum ity able to host irradiation positions inside, the capability of fre- Transmutation Experimental Facility’’, Grant Agreement No. FP7-232527. [3] Mercatali L, Serikov A, Baeten P, Uyttenhove W, Lafuente A, Teles P. Design quent fuel shuffling and fresh/spent fuel replacement in the core study of a fast spectrum zero-power reactor dedicated to source driven sub- is a desirable feature. For this reason the decay heat power after critical experiments. Energy Convers Manage 2010;51:1818–25. a possible real cycle composed by five sub-cycles was evaluated, [4] Rubbia C et al. 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