THICKET 2008 – Session III – Paper 05

The VVER Code Validation Matrix and VVER Specificities

Ivan Tóth

KFKI Atomic Energy Research Institute, 1121 Budapest, Konkoly-Th. u. 29-33, Hungary

Abstract. CSNI activities to support the safety of the Russian VVER-type reactors are summarised. The most important action was the development of the VVER-specific Code Validation Matrix (CVM) as a supplement to the earlier CSNI CVM for PWRs. Objectives and structure of the CVMs, along with VVER- specific phenomena are described and an overview of selected test facilities and tests is given. Section 3 presents the VVER-related OECD actions: the PSB, Bubbler-Condenser and Paks Fuel projects. Among CSNI’s International Standard Problems (ISP) only one was devoted to VVERs: ISP33 based on the PACTEL facility. Therefore also the earlier IAEA activities in this field are reviewed, with the four Standard Problem Exercises (SPE) based on the PMK test facility. The tests and outcome of the computer code analyses are described. Although not a CSNI action, major conclusions of a series of seminars on horizontal steam generators are also summarised.

1. INTRODUCTION

CSNI activities in the field of VVER reactors of Russian design started in the early nineties by forming a Support Group with the mandate to review and collect VVER-specific test facilities and tests in order to support validation of thermal-hydraulic codes. After several years of work the results could be summarised in a report [1].

As a follow-up of this activity, CSNI first decided to support the final stages of construction of the PSB integral-type test facility located in Electrogorsk, Russia and then organised an international project with the aim to conduct tests at PSB that would address phenomena not yet covered in the VVER-specific CVM.

In the long series of CSNI ISPs only one, ISP33 [2], was devoted to VVERs: the test was conducted at the Finnish PACTEL facility and addressed the natural circulation behaviour in single- and two-phase conditions.

Although directly not supported by CSNI, it is worth-while to mention two activities, where – besides VVER- operating countries – several OECD countries were also involved. Under the umbrella of the IAEA four SPEs were conducted in the period 1985-94 [3-6] based on tests performed at the PMK facility in Hungary. More than 20 countries from all over the world participated in the code validation activity.

A series of international seminars was initiated in 1991 by Finland, the sixth of these series being held in these days in Russia. The seminars are devoted to specific issues of horizontal steam generators – as applied in VVER-type NPPs – and address items, like experiments, modelling and structural aspects of such steam generators.

2. THE VVER CODE VALIDATION MATRIX

2.1 Objectives

In the early nineties an OECD Support Group was created with the mandate to review the level of validation of advanced thermal hydraulic codes applied for the analysis of VVER reactor primary systems in accident and transient conditions. The aim was to develop a supplement to the existing ITF and SETF CCVMs [7, 8] under consideration of the specific features of VVER reactor systems and their behaviour in normal and abnormal situations. This includes the necessary enlargement of the experimental data base for code assessment with data which were not taken into account in the previous CSNI CCVMs. At present, it is limited to large and small break LOCAs and transients and does not include shutdown transients and accident management scenarios. 61 Ivan Tóth (KFKI, Hungary)

The objectives of the OECD Support Group were:

• to identify the phenomena relevant in VVER reactor primary and secondary systems during LOCAs and transients. • to compare the phenomena of VVER reactor systems with LWR reactor systems and to clarify similarities. • to describe the phenomena involved in details as the basis for a common evaluation and an assessment by experimental data. • to identify test facilities and experiments that supplement the CSNI CCVMs and are suitable for VVER specific code assessment. • to establish criteria for the quality requirements and completeness of data finally to be used for the VVER specific code validation.

The activity included the following steps:

1. Characterisation of the main features of VVER reactor systems that are relevant to the thermalhydraulic design and the safety evaluation. Emphasis was given to hardware and operational features that distinguish VVER from Western PWR-systems. 2. Description of postulated accident scenarios. Again the main effort was to characterize thermalhydraulic aspects and ranges of parameters distinguishing VVER from Western PWR-systems. 3. Identification of facilities and of experiments that supplement the PWR and the BWR SETF and ITF CCVM and are suitable for code assessment. 4. Establishment of the VVER validation matrices that basically include range of conditions already covered by the previous available CSNI matrices, but also include VVER specific phenomena and features.

2.2 Specific Features of VVERs

From the hardware point of view the main differences between Western PWR-systems and VVER-systems are the following:

VVER-440 • six loops of primary circuit, • loop seals in hot legs, • horizontal steam generator with two headers, • elevation of the top of steam generators tubes related to the top of the active core (about 4 m, PWR about 10 m), • shrouded fuel assemblies with hexagonal fuel rod arrangement, • injection points of ECCS, • secondary side water volume in steam generators compared with nominal thermal core power is larger, • two isolation valves in each primary loop, • special pressure suppression system (bubble condenser), • each control rod consists of two parts: lower fuel assembly and upper absorber, • lower plenum volume larger.

VVER-1000 • horizontal steam generators with 2 headers, • ECCS injection points, • secondary side water volume of the steam generators compared with the nominal thermal core power is larger, • lower plenum internal structures, • fuel assemblies with hexagonal fuel rod arrangements.

From operational point of view differences are present in relation to: • operational conditions and set points of actuation of ECCS, • working conditions of secondary side of steam generators and set points for the operation of feedwater and steam line.

The considered differences may lead to different phenomena or may affect the course of the transients. As an example, natural circulation between core and steam generators may be affected by the difference in the elevations. The presence of the hot leg loop seals may prevent the ’reflux condensing’ natural circulation mode in VVER-440.

62 THICKET 2008 – Session III – Paper 05

Additionally, different condensation rates can be expected inside the upper plenum when accumulators are actuated. Differences in secondary side water volume of steam generators may cause different transient evolution e.g. in a Loss of Feedwater transient.

2.3 Structure of the Cross Reference Matrices

Cross Reference Matrices related to LOCA and Transients were drawn up with the objective of allowing a systematic selection of tests suitable for code assessment. Table 1 presents, as an example, the CRM for large break LOCA.

Each matrix is composed of six sub-matrices related to the following items: • phenomena covered by CSNI matrix, • phenomena versus plant types, • phenomena versus test types, • test facilities versus phenomena (both system and separate effects tests), • test types versus test facilities (only system tests), • plant type versus test facility.

In the term ‘phenomena’ all the important thermal-hydraulic processes expected to occur during an accident are included. In the column “CSNI” it is indicated whether a phenomenon has already been evaluated by experimental investigation within the CSNI frame, “plant type” gives a ranking according to the characteristics of VVER-systems, “type of test” relates to the definition of the experiment, the meaning of “test facilities” is self evident, both system test (integral facilities) and separate effects facilities are included.

Principles of the phenomena selection will be discussed in section 4. Test types and test facilities were selected essentially on the basis of personal experience of the participants of the Support Group, and, more specifically, on the basis of the knowledge of the national representative to which the test facility belonged. However, criteria for selection were commonly agreed upon, such as: • general technical suitability, • experimental coverage of phenomena,

Table 1: Cross Reference Matrix for Large Break LOCAs in VVERs.

Matrix I Plant TEST FACILITY *1 CROSS REFERENCE MATRIX FOR Test type type LARGE BREAKS System Tests Separate Effects Tests

- Test facility vs phenomenon - CSNI + suitable for code assessment + covered by o limited suitability o partially covered - not suitable - not covered x expected to be suitable

- Phenomenon vs plant type - Test type vs test facility + fully specific to WWER + already performed o partially specific o performed but of limited use - not specific - not performed

- Phenomenon vs test type - Plant type vs test facility + occuring + covered by o partially occuring o partially covered - not in list - not covered CSNI WWER-440/213 WWER-1000 Blowdown Reflood Refill PSB-WWER PM-5 SB ISB-WWER bank (EREC) Data GWP REWET-II IVO-CCFL SKN SVD-1 SVD-2 TVC-440 EVTUS KS TOPAZ SG-NPP FLORESTAN Break flow +--+++x +xx Phase separation o+oo++x x o Mixing and condensation during injection o++o++x x Steam condensation in SG primary side -oo---x x + 2-phase flow in SG primary and secondary side - ooooox x Core wide void + flow distribution o++o++xo - ECC bypass and penetration o++oo+- - CCFL (UCSP) o++o++x+ x + Steam binding (liquid carry over, ect.) ooo-+ox +x Pool formation in UP o++- ++x+ x Phenomena Core heat transfer incl. DNB, dryout, RNB o+++++xo+x + +++oo+ + Quench front propagation o++o+ox +x ++ +x+o + + Entrainment (Core, UP) -+oo+oxo+x + xo Deentrainment (Core, UP) -+o-+oxo+x + xo 1 - and 2-phase pump behaviour -++ +00x - Noncondensible gas effects -oo-++x x x ISB-WWER o+ ---important test parameter *1 refer to description of test facilities PSB-WWER + ---- leak location/leak size SB +++++ - pumps off/pumps on - ECC injection mode PM-5 + + Test Facility Test

System Tests System - ACC-pressure

63 Ivan Tóth (KFKI, Hungary)

• adequacy of instrumentation, • adequacy of documentation.

Since the aim of the Support Group was to review all test facilities which fulfilled the above criteria, no pre- selection was made with respect to availability of the data. The list of test facilities given in Appendix D of the report can be considered as an exhaustive one, from which tests for code validation purposes can be selected. The main emphasis was laid on integral systems, but a large number of separate effect test facilities was also included.

The symbol X, introduced for these matrices (but not employed in the CSNI matrices), shows that the new facilities, are especially suitable for the simulation of identified issues not covered in previous facilities.

2.4 Relevant VVER-Specific Phenomena

2.4.1 Basis for phenomena selection For the selection of the phenomena three principles were applied: • The first principle is that the phenomena identified in the CSNI matrices are in general also relevant to VVERs because of common characteristics of PWR and VVER-systems. Therefore it is important to stress that code validation and assessment plans for thermalhydraulic codes to be used for safety assessments of VVERs should be made on the basis of both: the ITF CCVM and SETF CCVM as well as on the VVER-specific matrices. • The second principle for selection of the phenomena for the VVER matrix is their relevance to safety. The selected phenomena have to be important to safety and furthermore their accurate modelling in computer codes crucial to safety analyses. A section of the report provides a tabular overview of the selected phenomena and an appendix gives a detailed description of the phenomena and discusses their safety relevance. • The third principle for selection of phenomena relates to accident scenarios. The phenomena were identified for three separate accident scenario groups and for these separate cross-reference matrices were developed. These groups are large break LOCA, small and intermediate break LOCA and transients. Other scenarios, in particular shutdown and accident management transients should be considered in a future revision of the report.

2.4.2 Phenomena description – an example The selected phenomena were thoroughly described and their relevance to safety discussed. As an example the description of the loop seal clearing phenomenon – as described in [1] – is reproduced below.

Description of the phenomenon of loop seal clearing (cold leg) In most current VVER designs, the primary piping from the outlet of the steam generators descends to a level below the mid level of the core before rising to enter the pump inlet from below. During a LOCA, when the primary circuit is partially voided, liquid present in the U- bend of this intermediate leg can form a barrier to the flow of steam around a loop. This liquid plug is said to form a loop seal. A pressure difference across the loop seal, which is greater than the hydrostatic pressure arising from the height of the pump inlet side of the intermediate leg is necessary before steam can displace the loop seal.

The manner and extent of the displacement of the liquid from the loop seals depend on the steam supply characteristics. At reduced steam velocities and low liquid subcoolings, the expulsion of fluid will be incomplete. This aspect of loop seal behaviour is closely connected with the phenomenon of counter-current flow limitation (CCFL). However the pipe sizes of interest are very much larger than those studied in most separate effect tests.

When the loop seals clear, their liquid contents are displaced into the reactor vessel, producing a rapid refill. The number of loop seals clearing, which depends for instance on the size of the break, affects the total amount of water added. How many loops need to clear is determined by the balance between the frictional pressure drop around the cleared loops and the hydrostatic pressure difference corresponding to the loop seal. Loop seals may be refilled after they have cleared. Back flow of ECCS water from the cold legs through the pump is limited by the height of the impeller and the diffuser geometry, but can occur if there is a general flow reversal, or if the level in the cold leg is high enough. Condensate from the steam generator tubes can also contribute to the refill of the seals.

Relevance to Nuclear Reactor Safety The loop seals have a particularly significant effect on the behaviour of a plant in a small cold leg break LOCA. Until at least one of the loop seals has cleared, steam produced in the core cannot reach the break and the pressure in the primary system remains high, at a level determined by the secondary side pressure. Furthermore, as the primary mass inventory decreases, the hydrostatic head of the liquid trapped in the pump inlet causes the liquid level in the core region to be depressed by an amount equal to the depth of the loop seal. This can lead to core uncovery. This phenomenon is similar to that in PWRs.

64 THICKET 2008 – Session III – Paper 05

2.5 Facilities and Tests

The test facilities listed in the report were selected irrespectively of the fact, whether the facility owners were ready to supply test data to a data bank or not.

The first set of test data was selected along the principles that they • were performed on well-known test facilities, • cover a wide range of VVER-specific phenomena, • are of high quality.

Criteria for facility and test selection were identified, including guidelines to qualify both facilities and tests. It is important to note that there is a close interconnection between facilities and tests. Therefore, the qualification of facilities and tests performed in that facility were made in parallel. For the selection of tests suitable for computer code assessment the following requirements were considered: • representativity of the test with regard to the reactor conditions including the range of parameters, • quality of the data measured with adequate instrumentation and with acceptable uncertainties, • quality and completeness of test documentation, • scaling considerations and boundary conditions.

As a result of the qualification process the facilities and tests given in Table 2 were selected.

2.6 Validation Matrices

In Chapter 3 the Cross Reference Matrices (CRM) were presented. In the previous Chapter the selected facilities and tests were described. In order to support code validation the so-called Validation Matrices (VM) have to be constructed. The VM represents two submatrices of the CRM: • the phenomena versus test facility and • the test type versus test facility submatrices, with the difference that data are presented here only for the selected tests and no more for the test facilities in general.

Table 3 gives an example of the VM.

Each test is evaluated against the phenomena on three levels: • “simulated”, meaning that the phenomenon occurs in the test and is representative for the real plant conditions, • “partially simulated”, when only some aspects of the phenomenon are occurring, or the simulation is only partially representative for plant conditions (e.g. due to scaling), • “not simulated”, with the obvious meaning that the phenomenon is not present in the test.

As a further information for code validation a separate submatrix indicates, which test type is addressed by which test (the test type addressed being marked by a “x”).

Table 2: Facilities and Tests selected for VVER Validation Matrix.

Facility Test Brief Description SB, EDO, Russia SB/1 100% Double break in cold leg SB, EDO, Russia SB/2 1% Cold leg break BD, EDO, Russia BD-1 Boron dilution SVD-2, IPPE, Russia 2 Dryout at low pressure PM-5, IPPE, Russia 6 Loop seal clearance KS, RRC-KI, Russia KS/19R/TF84 DNB, dryout KS-1, RRC-KI, Russia KS-1/05-91/N034 Heat transfer in covered and partially covered core ISB-WWER, EREC, Russia UPB-2.4 2.4% Upper plenum break ISB-WWER, EREC, Russia NC Natural circulation GWL, Skoda, Czech Rep. DNB-D DNB in 19-rod bundle REWET-II, VTT Energy, Finland SGI/9 Reflood PACTEL, VTT Energy, Finland ITE06 Natural circulation PACTEL, VTT Energy, Finland LOF-01 Loss of feedwater PMK-2, KFKI-AEKI, Hungary IAEA-SPE-4 CLB with secondary bleed and feed

65 Ivan Tóth (KFKI, Hungary)

Table 3: CRM for Small-Break LOCA.

Test facility PM-5 PACTEL PACTEL PMK-2 Phenomena IAEA- addressed by Test number 6 ITE-6 LOF-1 SPE-4 1 Natural circulation in 1-phase flow, primary side - + ? - 2 Natural circulation in 2-phase flow, primary side + + - + 3 Reflux condenser mode and CCFL - - - - 4 Asymmetric loop behaviour - + - - 5 Leak flow o - - o 6 Phase separation without mixture level formation o o - o 7 Mixture level and entrainment in SG (SS+PS) - o o - 8 Mixture level and entrainment in the core o o - o 9 Stratification in horizontal pipes - o - - 10 ECC-mixing and condensation - - - o 11 Loop seal clearance (CL) - - - o 12 Pool formation in UP/CCFL (UCSP) - - - - 13 Core wide void and flow distribution - o - - 14 Heat transfer in covered core + + + + 15 Heat transfer in partly uncovered core - o - + 16 Heat transfer in SG primary side o o o o 17 Heat transfer in SB secondary side - - o - 18 Pressuriser thermohydraulics - - - - 19 Surge line hydraulics - - - - 20 1- and 2-phase pump behaviour - - - - 21 Structural heat and heat losses - - - - 22 Noncondensible gas effects - - - - 23 Phase separation in T-junct., and effect on leakflow - - - - 24 Nat. circul., core-gap-downcomer, dummy elem. - - - - 25 Loop seal behaviour in HL o + - o 26 Recirculation in the SG primary side - o - - 27 Boron mixing and transport - - - 28 Water accumulation in SG tubes - - -

Test versus phenomena + Simulated o Partially simulated - Not simulated

IAEA- Test Types 6 ITE-06 LOF-1 SPE-4 1 Stationary Test addressing energy transport on primary side x 2 Stationary Test addressing energy transport on secondary side x 3 Small leak overfeed by HPIS, secondary side necessary 4 Small leak without HPIS overfeeding, secondary side necessary x 5 Intermediate leak, secondary side not necessary 6 Pressuriser leak 7 SG-tube rupture 8 SG-header rupture

2.7 Further Activities

As it can be seen from the validation matrix presented there is a number of phenomena which have not been adequately addressed by available experimental programs or tests. Due to the general similarity between VVERs and PWRs for most of these phenomena, e.g. for 3D effects, integral ECCS behaviour, number of transient phenomena and flow maps in pipes, the code validation may be sufficiently done using the CSNI ITF and SETF validation matrices. For the phenomena, which are VVER-specific, adequate experimental tests should be proposed and carried out. This is in particular true for large-scale integral simulation of the VVER system behaviour during large break LOCA accidents (including large leak from primary to secondary circuit). CSNI took an engagement in this respect, by proposing and supporting the international PSB project for VVER-1000 reactors, as discussed below.

A periodic updating of the matrices is necessary to include new relevant experimental facilities and tests (e.g. investigating boron dilution or behaviour of advanced reactors).

66 THICKET 2008 – Session III – Paper 05

3. VVER-RELATED OECD PROJECTS

3.1 The PSB Project

3.1.1 Main characteristics of the PSB facility The PSB-VVER test facility is a four loop, full pressure scaled down model of the primary system of the NPP with VVER-1000 reactor. Volume–power scale is 1:300, while elevation scale is 1:1.

It consists of a pressure vessel with an electrically heated rod bundle which provides full power conditions and has four identical loops with all major components and safety features of the reference VVER-1000 reactor. Among others, it simulates steam generators with horizontal tube arrangements. The PSB-VVER test facility is at present the only large-scale test facility suited for the investigation of safety-relevant thermal-hydraulic phenomena that are anticipated to occur in postulated accidental conditions for VVER-1000 reactors.

A PSB-VVER isometric view is presented in Fig. 1 and its main parameters are listed in Table 4. All the main components of the primary system of the VVER-1000 are simulated in PSB-VVER.

PSB-VVER instrumentation The PSB-VVER test facility is equipped with 1000 measurement channels including standard and non-standard devices. The following parameters are measured:

Upper plenum SG 2 SG 3

SG 4 Pressurizer SG 1

Pump 2 Pump 3

Pump 1

Pump 4

Downcomer

Core by - p ass Core model

Figure 1: Schematic View of the PSB-VVER Test Facility.

Table 4: The main PSB-VVER design parameters

Primary system Pressure 20 MPa Fluid temperature 350 °C Rod bundle electrical power 10 MW Coolant flow rate in the core 280 m3/h Core by-pass electrical power 80 kW Pressuriser heater power 80 kW Secondary system Pressure 13 MPа Steam temperature 320 °C Feed water temperature 270 °C Feed water flow rate for one SG 5 m3/h 67 Ivan Tóth (KFKI, Hungary)

• pressure • differential pressure • coolant temperature • wall temperature • mass flow rate (single and two phase flow) • flow velocity • local void fraction • collapsed level • heat flux • electrical power • momentum flux • coolant density

3.1.2 Aims and results of the project The project is devoted to the organization and implementation of experiments to study VVER-1000 accidental thermal-hydraulics events, by making use of the unique features of the large-scale integral test facility PSB-VVER- The selection of tests was made in accordance with the needs that emerged from the OECD/NEA validation matrix developed for VVER application. Pre- and post-test analyses with best-estimate system thermal-hydraulic codes have accompanied the tests. The project aims at generating new high quality test data, covering “white spots” in VVER validation matrix, so that best-estimate thermal-hydraulic codes can be validated for VVER application.

Despite the fact that the basic physics and thermal-hydraulics of VVER and PWRs of Western design are essentially similar, the design characteristics of VVER reactors (e.g. large primary and secondary side fluid inventory, horizontal steam generators, core design) entail specific types of accident sequences. As such, it is generally recognised that the safety evaluation of VVERs requires dedicated experimental and analytical analyses in order to assess the performance of safety systems and the effectiveness of accident management strategies. Clearly, the predictive capabilities of computer codes used in reactor safety analysis should be validated against relevant test data.

It is generally recognised, however, that the reference experimental database relevant to accidents and transients for VVER-1000 reactors is rather limited, especially for integral tests. According to the OECD VVER-specific code validation matrix [1], there is a broad area where VVER-1000 integral test data are practically absent.

The experimental programme covers conditions that are highly relevant for code validation related to PWR in general and for VVER-1000 safety assessments in particular. The test matrix has been defined with input from the Russian Safety Authority Gosatomnadzor, and is shown in Table 5.

Most of the tests have been completed according to the project schedule. The last test was considerably delayed and finally the decision was taken to perform it first at reduced power. This additional test, Test #5a was successfully performed on 24 January 2008 at 1.5 MW (i.e., 15%) power. Analysis of the test indicates that an adequate simulation of the large break accident (a guillotine cold leg break) can be realized in the PSB-VVER integral test facility. Regarding the Test #5, EREC currently plans its performance by June 2008. In the mean time most of the participants are performing post-test calculations of Test 5a (15% power) and pre-test calculations for Test 5 (100% power). A final project meeting is planned for September 2008 with the aim to discuss Test 5 results and post-test calculations.

3.2 The Bubbler-Condenser Project

The VVER-440/213 pressurised water reactors are fitted with a pressure-suppression system, called bubbler- condenser (Fig. 2), having the function to reduce the containment pressure in case of a design basis accident (DBA), such as a loss of coolant accident (LOCA). The device consists of 12 vertically distributed trays (Fig. 3) communicating with each other and they are located in a tower connected to the reactor containment building. Each tray is flooded with a pool of cold water (at room temperature) and includes gap-cap inlet openings. In the unlikely case of a LOCA, the steam from the primary circuit of the reactor and air enter the bubbler-condenser tower and are forced by the gap-cap system to bubble into the cold water present at each floor of the bubbler-condenser. This causes the steam to condense, thus maintaining both temperature and pressure within containment below given limits during the entire course of a postulated design basis accident.

The early works related to the BC and performed in the former USSR have been summarised in a status report by OECD NEA CSNI [25]. The first action within OECD was to establish a Support Group on "VVER-440 Bubbler

68 THICKET 2008 – Session III – Paper 05

Table 5: Experiments of the OECD-PSB-VVER Project.

No ID Test Phenomena observed Justification Note (date) (expected) 1 UP-11-08 11% Upper Two- phase flow NC, primary side Identification of Counter-part of the (2002). Plenum Asymmetrical loop behavior similarities and ISB-VVER test break” Leak flow differences in Separation of phases without mixture level processes / phenomena formation behavior at ISB- Mixture level and entrainment in SG VVER and PSB- Mixture level and entrainment in the core VVER (i.e. Stratification of flow in horizontal pipes manifestation of Loop seal clearance in the cold legs “scaling effect”) and Pool formation in the upper plenum their quantitative Heat transfer in the covered core characterization Heat transfer in a partially uncovered core Thermal - hydraulics in pressurizer Effects of the integral system 2 NC-1 Natural Single-phase natural circulation Basic phenomena in Parametric study (2003) Circulation Two-phase natural circulation transients, need of data Heat transfer on the SG primary side in “clean” boundary Heat transfer on the SG secondary side conditions (outcome Heat transfer in covered core of VVER Validation Heat transfer in partly uncovered core Matrix). Loop seal clearance Recommended by Asymmetric loop behavior Gosatomnadzor Structural heat and heat losses 3 CL-4.1- 4.1% cold Natural circulation of two-phase flow Relevant bounding Counter-part of 03 leg break Asymmetrical loop behavior case phenomenon. LOBI, BETHSY, (2004) test Separation of phases without mixture level Data lacking, as from SPES and LSTF formation recommendation of Mixture level and entrainment into SG OECD Support Group Mixture level and entrainment into the core on the VVER Loop seal clearance in the cold leg Validation Matrix. Pool formation in the upper plenum Relevant for VVER Heat transfer in the covered core safety assessment. Heat transfer in a partially uncovered core Recommended by PRZ thermal-hydraulics Gosatomnadzor Surge line hydraulics Structure heat and heat losses Effects of the integral system 4 PSh-1.4- Primary Two-phase natural circulation Absence of test data Analytical Exercise 04 to Asymmetrical loop behavior for “primary to (Blind test) (2005) secondary Leak flow secondary leak” leak Mixture level and entrainment into SG (VVER Validation Mixture level and entrainment into the core Matrix). ECCS mixing and condensation Recommended by Loop seal clearance in the cold leg Gosatomnadzor. Pool formation in the upper plenum Highly relevant for Heat transfer in the covered core VVER safety Heat transfer in a partially uncovered core assessment from point Heat transfer on the SG primary side of view of radiological Heat transfer on the SG secondary side consequence PRZ thermohydraulic Surge line hydraulics Structure heat and heat losses Effects of the integral system 5a - Large cold (Phase separation) Important for safety Test not planned (2008) leg break (Mixing and condensation during injection) assessment originally @ Low (Two-phase flow in SG primary and evaluations, strongly Post test analyses in Power secondary sides) supported by progress by several (15%) (CCFL) Gosatomnadzor Institutions using: (Pool formation in UP) RELAP5, RELAP5- (Core heat transfer) 3D, CATHARE and ATHLET, APROS 5 - Large cold (Phase separation) Important for safety Full power leg break (Mixing and condensation during injection) assessment experiment (10MW @ Full (Two-phase flow in SG primary and evaluations, strongly core bundle will be Power secondary sides) supported by installed in the (CCFL) Gosatomnadzor facility) (Pool formation in UP) TO BE EXECUTED (Core heat transfer) BY JUNE 2008

69 Ivan Tóth (KFKI, Hungary)

Condenser Containment Research Work" (1991-1994). The SG stated that supplementary research work is needed in various areas such as:

• dynamic loading of the cap-gap systems by mass flow induced differential pressure upon occurrence of a LOCA, • oscillatory loading of the flat water pool trays by condensation phenomena, • water carry-over into the air traps during the impulsive air transfer period.

The European Union launched a PHARE project PH2.13/95 “Bubbler-condenser Experimental Qualification” (BCEQ) in 1995. The main objectives of the project were to investigate experimentally and analytically the behaviour of the bubbler-condenser in DBA conditions. The following tasks were specified:

Figure 2. Containment System of VVER-440/213 (from [28]).

Figure 3: Details of the Bubbler-Condenser System (from [28]).

70 THICKET 2008 – Session III – Paper 05

• To design and to erect a test facility, which replicates a portion of a prototypical BC configuration of the Paks NPP; to perform thermal-hydraulic and fluid-structure interaction tests on a test prototype configuration which replicates the BC of Paks NPP; to perform pre-test and post-test analyses with appropriate computer codes. • To design and to erect a test facility which replicates a portion of a prototypical BC configuration of the Dukovany and Bohunice NPPs; to perform structural verification tests on the weaker pressure-retaining steel structure of Dukovany and Bohunice NPPs under differential pressure which occurs during the first moments of a LOCA; to perform pre-test and post-test analyses with appropriate computer codes.

Both the above tasks were accompanied by analytical support with established computer codes, in order to make sure that the test configurations reproduce NPP conditions and that the test results are relevant for the plants. In addition, some small-scale separate effect tests were defined to support the large-scale ones.

Within the PHARE project the bubbler-condenser test facility was constructed at EREC and three tests were performed. The BCEQ project came to the conclusions that the tests demonstrated the bubbler-condenser functionality and the physical parameters are far below the values which could create any risk in a DBA [26].

The OECD Support Group, after careful review of the interpretation and extrapolation of the PHARE project results recommended [27]:

• to perform further post-test analyses of the results obtained so far (incl. investigations of non-uniformities in temperature and flow velocity distributions observed in the EREC tests); • to use post-test calculation results for the bubbler condenser design qualification, code validation and modelling improvements; • to perform tests for completion of the EREC test matrix simulating MSLB, medium and small break LOCA accidents; • to perform further bubbler condenser investigations for the Kola nuclear power plant.

In 2001, in response to a request of their safety authorities the Hungarian, Czech and Slovak utilities took the initiative to perform a joint experimental programme based on the EREC test facility. Parallel to the initiative for establishing this consortium, the Hungarian Safety Authority (HAEA) requested the assistance of the OECD NEA CSNI for the preparatory phase of the experimental work as well as for the analyses of code calculations and experimental results. The CSNI supported the HAEA request and approved the establishment of a Bubbler- Condenser Steering Group with the following mandate:

• to produce convincing evidence that the VVER-440/213 type bubbler-condenser works during DBAs as designed • to help in the planning of the new EREC experiments and in the interpretation of the results • to provide well qualified experimental results serving as basis for the validation of best estimate calculation tools.

The EREC experiments and the related pre- and post-test calculations addressed the following postulated events:

• Main steam-line break (MSLB) of the Paks NPP. Pre- and post-test calculations were co-ordinated by the Paks NPP (calculated by VEIKI, Budapest). • Medium break in the cold leg of the Dukovany NPP with a diameter of 200 mm (MBLOCA). Pre- and post-test calculations were co-ordinated by the Dukovany NPP (calculated by NRI, Rez). • Small/medium break loss-of-coolant accident at the Mochovce NPP, in loop 1 with the pressurizer (90 mm). Pre- and post-test calculations were co-ordinated by the Mochovce and Bochunice NPPs (calculated by VUJE, Trnava). • The project came up with the final conclusions:

• The test parameters measured by different transducers provide values that are generally consistent with each other. • The discrepancies between the measured and calculated values are not significant and the calculations are conservative. • The observed differences between the measured and calculated values can be adequately explained. • The maximum pressure experienced in the tests is far from the 0.25 MPa design pressure of the containment system.

71 Ivan Tóth (KFKI, Hungary)

• The maximum pressure load on the tray walls measured during the tests, is far less than the 30 kPa limit value. • Water level fluctuations were experienced but were found to be minor and disappeared when the steam started to flow into the bubbler condenser pool. • Within the range of conditions explored in the EREC tests, condensation-oscillation phenomena were not observed. • The sequences investigated in the tests do not cause any significant challenge for the VVER-440/213 type bubbler-condenser and localization system.

3.3 The OECD-IAEA Paks Fuel Project

In 2003, during a fuel crud removal operation at the Paks-2 unit several fuel assemblies were severely damaged. The assemblies were being cleaned in a special tank in a service pit connected to the spent fuel storage pool. The first sign of fuel failure was the detection of fission gas release from the cleaning tank. Later visual inspection revealed that most of the thirty fuel assemblies suffered heavy oxidation and fragmentation. The first evaluation of the event showed that the severe fuel damage was due to inadequate cooling that first led to gradual heat-up of the coolant inside the tank and to steam formation and accumulation in the upper part in a couple of hours (Fig. 4).

The Paks-2 event was discussed in various committees of the OECD Nuclear Energy Agency (OECD/NEA) and of the International Atomic Energy Agency (IAEA). Recommendations were made to undertake actions to improve the understanding of the incident sequence and of the consequence this had on the fuel. It was considered that the Paks- 2 event may constitute a useful case for a comparative exercise of computer codes, in particular for models devised to predict fuel damage and potential releases under abnormal cooling conditions.

The OECD-IAEA Paks Fuel Project was established in 2005 as a joint project between the IAEA and the OECD/NEA. Thirty organizations from sixteen countries participated in the project. The participants performed analyses of the event using numerical models based on a common database.

Figure 5 shows that there is a factor 3 difference among the calculated inlet flow rates to the follower assembly, but all the calculations predict correctly the decreasing trend during the subcooled period, which is the result of the continuously changing elevation head balance between the regions inside and outside the assemblies. For the differences in the calculated flow rates the following reasons can be cited: different by-pass areas and loss- coefficients used by the participants, but also the models of friction losses in the laminar-turbulent transition region may be responsible. In view of the very delicate balance between elevation head losses and laminar flow friction losses the range of the results is acceptable. The figure also displays the outlet temperatures of assembly 12. The differences among the participants during the heat-up phase to saturation can be explained partly by the fact that the flow rates along the heated parts are rather different, but to some extent differences in assembly group powers and fuel heat capacities may be the cause as well.

The project came to the following conclusions with respect to the adequacy of computer code performance [29]:

¾ The simulations covering thermal hydraulic, fuel behaviour and activity release aspects captured well major events of the incident: • the timing of water level decrease in the cleaning tank, • the timing of first fuel failures that very probably took place as a result of ballooning and burst of fuel rods at high temperature, • the rate of released activity from the fuel.

Figure 4: Cleaning Tank Configuration and Heat-Up (from [29]).

72 THICKET 2008 – Session III – Paper 05

0.07 1600 VTT VTT GRS 1400 GRS 0.06 AEKI.1 AEKI.2 AEKI.2 BME VEIKI 1200 VEIKI KI/IRSN 0.05 VUJE KI/IRSN USNRC 1000 IVS VUJE 0.04 800 USNRC

600 0.03 400

0.02 200

0.01 Outlet temperature of assembly No. 12 [°C] 0 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 0 5000 10000 15000 20000 25000 30000 35000 40000 Inlet flowrate per assembly of assembly No. 12 [kg/s] Time [s] Time [s]

Figure 5: Variation of Inlet Flow Rate and Outlet Temperature of Fuel Assembly 12(from [29]).

¾ The numerical analysis improved the understanding of Paks-2 event and helped to precise some unknown parameters of the incident as e.g. • the by-pass flow at low flow rate amounted to 75-90 % of the inlet flow rate that led to the formation of steam volume, • the maximum temperature in the tank was between 1200-1400 ºC, • the degree of zirconium oxidation reached 4-12 %, • the mass of hydrogen produced was between 2.8 and 8.4 kg.

4. VVER-RELATED BENCHMARK EXERCISES

4.1 OECD ISP33

The only VVER-related International Standard Problem (ISP) performed under CSNI auspices was ISP33, based on the Finnish PACTEL facility [2]. PACTEL is a volumetrically scaled (1:305) integral test facility of the pressurised water reactors used in Finland, the VVER-440 type.

The peak operating pressures on the primary and secondary sides are 8 MPa and 4.6 MPa, respectively. The reactor vessel is simulated with a U-tube construction including separate downcomer and core sections. The core itself consists of 144 full-height, electrically heated fuel rod simulators with a chopped cosine axial power distribution and a maximum total power output of 1 MW, or 22% of scaled full power. The fuel rod pitch (12.2 mm) and diameter (9.1 mm) are identical to those of the reference reactor. The rods are divided into three roughly triangular-shaped parallel channels representing the intersection of the corners of three hexagonal VVER rod bundles.

Component heights and relative elevations correspond to those of the full-scale reactor to match the natural circulation pressure heads in the reference system. The hot and cold leg elevations of the reference plant have been reproduced, including the loop seals. The hot leg loop seals are a result of the steam generator locations, which are at roughly the same elevation as the hot leg connections to the upper plenum. The cold leg loop seals are formed by the elevation difference between the inlets and outlets of the reactor coolant pumps, just as in other PWRs.

For practical reasons, the total hot and cold leg pipe lengths of the reference plant were not duplicated in the PACTEL facility, being shortened by almost a factor or two. As a result, the pipe cross-sectional area was increased to duplicate the volume scaling factor used in the remainder of the facility and preserve the time scale for energy transport from the heat source to the sinks. As a consequence of the increased cross-sections the Froude number in the loop seals become closer to that of the full-scale plant.

Three coolant loops with double capacity steam generators are used to model six loops found in the reference power plant. The U-tube lengths and diameters in the PACTEL steam generators correspond to those of the full-scale models. The horizontal orientation of these steam generators is one of the distinguishing features of the VVER design. One consequence of this geometry is a reduced driving heat for natural circulation. Another notable feature is the relatively large secondary side water inventory, which tends to slow the progression of transients.

The facility includes a pressuriser, high and low pressure emergency core cooling systems, an accumulator and primary pumps. 73 Ivan Tóth (KFKI, Hungary)

ISP33 was a “double blind” ISP, i.e. not only the test results, but also the test facility behaviour was unknown to the participants in the first round. The ISP included a second, “open” phase, where calculation could be done knowing the experimental results.

The experiment was performed to observe the natural circulation flow behaviour under quasi-steady state conditions over a range of primary side inventory levels. Core power level of 3.4% (155 kW) scaled nominal power was selected for this test. In the experiment, the facility was heated until it approached the selected temperature and pressure and a steady state was established near these conditions. The primary coolant was drained from the lower plenum in several steps, allowing the system to restabilize for 900 seconds between each step. For the duration of the test, the secondary side conditions were maintained near the nominal full power operating conditions of the reference plant.

The effect of the first draining was rapid decrease of primary pressure until the saturation temperature reached the core outlet temperature. Almost all the water in the pressuriser burst into the hot leg of loop one. The fluid temperature in the hot leg decreased temporarily when the subcooled water slug from the pressuriser surge line reached the hot leg. Vapour started to accumulate at the top of the upper plenum. The flow in the downcomer was single-phase flow with a nearly constant flow rate.

During the second draining the amount of vapour in the upper plenum increased quickly. The loop and the downcomer mass flow rates increased due to the draining. When the swell level in the upper plenum fell below the hot leg pipe entrances, considerable voiding occurred also in the beginning of the hot legs. The flow rate dropped dramatically and became stagnant. The loop seals prevented the vapour to reach the steam generators. This flow stagnation was observed in each of the three loops simultaneously. The system pressure rose sharply when the flow rate dropped since energy transfer to the steam generators was interrupted. Fluid temperatures in the upper plenum, hot legs and at the core outlet followed the rising saturation temperature.

When the pressure rose, water was forced back into the pressuriser. After about two minutes of flow stagnation and a continued increase in the pressuriser water level, all three loop seals cleared and a surge of two-phase flow into the steam generators reduced the pressure. As the pressure dropped, a fraction of the water in the pressuriser returned to the loop and the flow stagnated again, which initiated another system pressure increase. This was repeated two additional times, with the peak water level in the pressuriser decreasing with each cycle.

During the next draining the primary pressure dropped quickly. Temperatures in the upper plenum followed the saturation temperatures. The pressuriser was depleted again and primary mass flow rates increased. When the draining stopped, a relatively steady two-phase flow was established between 3000 and 4000 s. The bulk of this flow was transported through one loop, though none of the loops were totally stagnant. The two loops with low flow rates had partially filled hot leg loop seals while the third loop seal was clear.

As the inventory was reduced further the water level in the two filled loop seals dropped and the flow rates continued to decline, finally becoming nearly stagnant. The heat transfer mechanism from primary to secondary changed to the boiler-condenser mode. Steam was condensed in the steam generators, collected in the cold legs, and returned to the core via the downcomer.

The downcomer flow rate was quite low because of the high energy transfer efficiency of this mode. Reflux condensation was not observed because of the steam generator and hot leg loop seal geometry. Condensation took place in the horizontal U-tubes and so there was no driving force for flow back towards the upper plenum. Moreover, it is apparent from the geometry how a steady flow of condensate back into the hot leg would eventually fill the loop seal and block the steam flow.

12 countries and 15 institutions using a variety of codes, participated in the exercise. The major findings of the ISP can be summarised as follows:

• The overall transient process was reasonably well predicted. • Main discrepancies concerned the predictions of flow stagnations and time of core heat-up. • 2-phase natural circulation flow rates were in general overpredicted. • Post-test analyses in general yielded improved results due to inclusion of some experimental problems not known in advance for “double-blind” predictions (e.g., leak of safety valve). • The loop seal behaviour remained a problem also for “open” post-test analyses.

74 THICKET 2008 – Session III – Paper 05

4.2 VVER-1000 Coolant Transient Benchmark

The NEA Nuclear Science Committee organised the VVER-1000 coolant transient (V1000CT) benchmark, which defines coupled code standard problems for validation of thermal-hydraulic system codes for application to Soviet- designed VVER-1000 reactors based on actual plant data. The overall objective is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in reactivity transients in a VVER-1000. The V1000CT benchmark consists of two phases:

• V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation • V1000CT-2 concerns the calculation of coolant mixing tests and main steam line break (MSLB) scenarios.

The reference problem chosen for simulation in Phase 1 is a MCP switching on when the other three main coolant pumps are in operation in a VVER-1000. It is based on an experiment that was conducted at the Kozloduy NPP Unit #6 as a part of the start-up tests. The test was carried out at 27.47% of the nominal power level with three of the four MCPs in operation. Due to one of the control rod groups partially inserted into the core the initial power distribution showed axial asymmetry. When at the beginning of the transient the MCP #3 is switched on, additional radial asymmetry is introduced by the colder water supplied to one quarter of the core. Simulation of the transient requires evaluation of the core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system. Three exercises are defined in the framework of Phase 1: a) Exercise 1 . Point kinetics plant simulation; b) Exercise 2 . Coupled 3-D neutronics/core thermal-hydraulics response evaluation; c) Exercise 3 . Best-estimate coupled 3-D core/plant system transient modelling. Phase 1 has been completed [30], [31], [32]. Results of the benchmark exercise will be presented in Paper 23.

Since earlier coupled code benchmarks indicated that further development of the mixing computation models in the integrated codes is necessary, a coolant mixing experiment and a MSLB scenario are selected for simulation in Phase 2 of the benchmark. As an option CFD modelling of the mixing process in the reactor vessel is also included in the exercise. For this specific case additional data from KNPP Unit #6 were made available. The selected mixing experiment was conducted by cooling and heating a single loop at 9 % of nominal power with all MCP in operation at the Kozloduy NPP Unit 6 as part of the plant commissioning phase.

The transient to be analyzed in Phase 2 is initiated by a MSLB in the VVER-1000 NPP between the steam generator and the steam isolation valve, outside the containment. The event is characterized by asymmetric cooling of the core, stuck rods and a large primary coolant flow variation. The main objective is to clarify the local 3-D feedback effects depending on the vessel mixing. Special emphasis is put on testing 3-D vessel thermal-hydraulic models and coupling of 3-D neutronics/vessel thermal-hydraulics. The MSLB scenario simulation is divided into two exercises: Exercise 2 consists of coupled 3-D neutronics/vessel thermal-hydraulics simulation using specified vessel thermal- hydraulic boundary conditions, and Exercise 3 consists of best-estimate coupled 3-D core/3-D vessel/plant system modelling.

4.3 IAEA Standard Problem Exercises

Recognizing the lack in code validation for VVER-type power plants Hungary proposed to IAEA in the early eighties to conduct Standard Problem Exercises (SPE). The freshly constructed PMK facility was offered as experimental basis for the exercises.

The PMK-2 facility is a scaled down model of the Paks NPP and it was primarily designed for investigating small- break loss of coolant accidents (SBLOCA) and transient processes of VVER-440/213 plants. The volume and power scaling are 1:2070. Transients can be started from nominal operating conditions. The ratio of elevations is 1:1 except for the lower plenum and pressuriser. The six loops of the plant are modelled by a single active loop. In the secondary side of the steam generator the steam/water volume ratio is maintained. The coolant is water under the same operating conditions as in the nuclear power plant.

The core model consists of 19 electrically heated rods, with uniform power distribution. Core length, elevation and flow area are the same as in the Paks NPP.

In the modelling of the steam generator primary side, the tube diameter, length and number were determined by the requirement of keeping the 1:2070 ratio of the product of the overall heat transfer coefficient and the equivalent heat

75 Ivan Tóth (KFKI, Hungary) transfer area. The elevations of tube rows and the axial surface distribution of tubes are the same as in the reference system. On the secondary side the water level and the steam to water volume ratios are kept. The temperature and pressure are the same as in the NPP.

Cold and hot legs are volume scaled and care was taken to reproduce the correct elevations of the loop seals in both the cold and the hot legs. Cold and hot leg cross section areas if modelled according to volume scaling principles would have produced much too high pressure drops. Since, for practical reasons, length could not be maintained 1:1, relatively large cross sections were chosen for the PMK-2 loop. On the one hand this results in smaller cold and hot leg frictional pressure drops than in the NPP, on the other hand, however, it improves the relatively high surface to volume ratio of the PMK-2 piping. As to the former effect, the small frictional pressure drop of the PMK-2 cold and hot legs will have a negligible effect on small-break processes. However, the pressure drop is increased using orifices around the loop.

For the pressuriser the volume scaling, the water to steam volume ratio and the elevation of the water level is kept. For practical reasons the diameter and length ratios cannot be realised. The pressuriser is connected to the same point of the hot leg as in the reference system. Electrical heaters are installed in the model and the provision of the spray cooling is similar to that of Paks NPP.

For the hydroaccumulators, the volume scaling and elevation is kept. They are connected to the downcomer and upper plenum similar to those of the reference system. The four hydroaccumulators of the Paks NPP are modelled by two vessels. The HPIS and LPIS systems are modelled by constant coolant flow rate.

The main circulating pump of the PMK-2 serves to produce the nominal operating conditions corresponding to that of the NPP prior to break initiation as well as to simulate the flow coast-down following pump trip early in the transient. For this reason the pump is accomodated in a by-pass line. Flow coast-down is modelled by closing a control valve in an appropriate manner and if flow rate is reduced to that of natural circulation, the valve in the by- passed cold leg part is opened while the pump line is simultaneously closed.

Four exercises were conducted in the period 1985 to 1994 [3-6]. Intermediate break LOCAs were selected for three of the SPEs (SPE-1, 2, and 4) with break size of 7.4% in the cold leg. All tests started from an initial condition representing the nominal operating parameters of the power plant. The three tests differed in the availability of ECCS:

• no hydroaccumulators and only one of three HPIS available (SPE-1); • three hydroaccumulators and one HPIS (SPE-2); • three hydroaccumulators and no HPIS, only LPIS representing a BDBA sequence, with secondary bleed and feed as AM procedure (SPE-4).

SPE-3 addressed a VVER-specific case, the opening of the lid of the steam generator hot side collector, resulting in a leak from the primary to the secondary side. In the test the availability of 3 hydroaccumulators and 2 HPIS was assumed.

The significance of the IAEA coordinated SPEs was that it convened a large number of participating organisations from all over the world. All in all 26 countries and 40 institutions participated in the exercises. From the series of PMK SPEs the following conclusions were drawn:

• The first two exercises provided unique opportunities for institutions not having important background in safety analysis to get experience in code validation, to learn lessons from more experienced countries. The level of knowledge was much more equalized in SPE-3 and SPE-4. • VVER data base became available for non-VVER countries for validation activities of their codes (e.g. CATHARE, ATHLET, RELAP5). • Sophisticated codes as RELAP5, CATHARE, ATHLET and DYNAMIKA have demonstrated their ability to model overall SBLOCA behaviour in VVER-440/213 systems. • The hot leg loop seal behaviour could be modelled sufficiently well, but this was not the case for the cold leg loop seal, especially for the clearing of it. It affects the coolant distribution in the loop. • Partial and temporary core uncovery appeared in SPE-1 and SPE-4, especially in the latter extended dryout occurred. Prediction of dryout proved to be difficult for most analysts. • Further efforts are needed to correctly model the horizontal steam generator. • Further improvements are needed for the correct modelling of the hydroaccumulator injection: most code results predicted stepwise injection in SPE-2, while it was continuous in the test. Also the distribution of injected accumulator water between vessel and loop caused difficulties.

76 THICKET 2008 – Session III – Paper 05

5. ISSUES OF HORIZONTAL STEAM GENERATORS

More stringent safety requirements have brought up an increased need for accurate modeling of horizontal steam generators, which are a characteristic feature of the VVER reactor type. Horizontal U-tubes in several layers and a large shell side water pool bring the applications of one-dimensional thermal-hydraulic codes under suspicion. That is why it is exceptionally important to verify such a use of the codes for horizontal steam generators. The lack of publicly available experimental data has hindered verification of the obtained calculation results.

The above considerations led to an initiative in Finland to organize an International Seminar, where these issues can be handled among experts. The participants of this first seminar in 1991 were invited mainly on the basis of bilateral relations from the USSR, Germany, Hungary, France, Poland and Finland. The objective of this seminar was defined as to collect all existing information and experience in the area of horizontal steam generator modeling for the VVER-reactors. More specifically the seminar was to cover the following subjects:

• compare existing models • define additional experimental needs • initiate necessary experimental activities • develop improved models for horizontal steam generators.

Concerning the first objectives the seminars pointed out that the calculation efforts should meet two clearly different objectives. First, the modeling capability of the system codes should be developed to such a level that would satisfy the need to calculate the steam generator behaviour during all conditions, including transient and accident conditions. Secondly, there is a need for more accurate flow calculations to predict the fluid flow conditions on the steam generator secondary side. To do this, 3-D calculation codes need to be developed. Such calculations were deemed to be of crucial importance for determining thermal hydraulic boundary conditions for the structural evaluations of any proposed design changes.

Applications of system codes RELAP5, ATHLET, CATHARE, APROS and DINAMIKA were presented [9-12]. The main issues of code application are whether the 1-D nodalisation model is sufficient, how detailed nodalisation is required, and how to obtain relevant data for validation and assessment. It was pointed out that the required detail of nodalisation depends on the problems to be solved. The safety analyst should understand the role and importance of the model from the viewpoint of the specific application. Too detailed models may become impractically large for the computer capacity. On the other hand, too simple models may give totally wrong picture of the process.

Already during the first seminar a proposal was put forward to organize a common calculation exercise for the analyses of secondary side void, temperature and flow conditions. First experimental data supplied by OKB Gidropress were not found by participants to have sufficient quality. After obtaining new data on the secondary side water inventory an exercise was proposed to calculate by various codes the secondary side water inventory e.g. on power levels 0%, 40%, 60%, 80% and 100%. Unfortunately, this activity was not pursued further in following seminars.

The experimental basis of the thermal-hydraulic behaviour is not yet sufficiently wide. PACTEL results with a new steam generator confirmed the theoretically predicted flow reversal in the lower tube bundle. The PMK-2 experiments with the better instrumented steam generators have also shown the phenomenon.

Tests from the PACTEL facility concerning heat transfer degradation and flow distributions during boil-off under natural circulation conditions yielded the following results [13]:

• Circulatory flow pattern in the tube bundle is characteristic to natural circulation conditions, with continuous flow from the hot to the cold collector in the upper part, dry or wetted, and reversed flow at the bundle bottom. • Primary to secondary heat transfer rate generally follows the wetted tube area. At higher secondary levels the power throughput tends to decrease slightly faster with the secondary level than the wetted area does; at lower levels (<50%) the recirculatory flow greatly enhances primary to secondary heat transfer. • Mixing of the reversed, cold flow in the hot collector is confined to inside and above the conical expansion piece. Its influence on the overall loop natural circulation is small.

Direct measurements of water inventory of the VVER-440 steam generator in operation (performed at Rovno Unit 1) were discussed by OKB Gidropress [14]. Thermal hydraulic experiments were presented by Mitsubishi Heavy Industry on the flow behaviour of the secondary side of the horizontal steam generator designed for the passive next generation PWR concept [15].

77 Ivan Tóth (KFKI, Hungary)

The effect of non-condensable gases on steam generator heat transfer and on integral system behaviour was investigated in PACTEL [16]. Tests conducted in two-phase conditions on the primary side did not show important differences depending on the gas applied (air and helium): the lowermost tubes were filled with gas and heat was transferred via the upper ones, where steam was condensed. In the single-phase test air occupied the top part of the tube bundle and only the lowest tubes took part in the heat transfer process. In all cases the primary pressure and hot leg temperature increased as a consequence of reduced heat transfer area. This behaviour was also shown by earlier PMK experiments [17].

The separate effect experiments directed to study the flow distribution in the feedwater distributor of PGV-440 were discussed [18]. The results help to explain the corrosion-erosion mechanism, which has been found to cause damage to the distributor. The low and unstable flow rate through the nozzles near the T-joint creates the possibility of steam bubbles entering the collector that enhances corrosion-erosion.

The lack of sufficient experimental data to support the development was recognized. Some help to the situation can be expected, since a steam generator data bank is being collected in Russia [19]. In the near future this data bank will be made available, and it should be useful for the model development and validation purposes.

Already during the first seminar it was suggested that the scope should be enlarged to structural problems with special respect to the safety of the hot collector.

Concerning the mechanical integrity aspects of the primary collectors, the significance of temperature stratification under natural circulation conditions was discussed and scoping experiments were presented [20]. Experiments visualizing the conditions in the hot and cold steam generator collectors indicate the formation of a stratification layer in both collectors under natural circulation conditions. This leads to important thermal stresses in the collector walls [21].

The acute problem of feedwater distributors experiencing severe erosion-corrosion problems received much attention in the seminars. Two principally different designs exist for replacing the eroded pipe sections. One is developed by Gidropress of Russia, and it has been installed into six steam generators in the Ukrainian Rovno plant, and into one at Loviisa Unit 2. The design employs the idea of locating the new distributor and of injecting the feedwater through the short nozzles above the tube bundle. The other design has been developed in Czech Republic by Vítkovice. Again the distributor locates above the tube bundle, but the feedwater is injected in the middle of the tube bundle through long nozzles. Such feedwater distribution systems have been installed into sixteen steam generators at Dukovany. A modified version with a mixing chamber has been installed into two steam generators at Bohunice Unit 3.

A large leakage accident from the primary-to-secondary side (PRISE) is possible in the horizontal VVER steam generators. The break can be a lift-up of the primary collector cover, or a ductile failure of the collector wall in the area of tubing connections. That is why the large PRISE accidents (the break area up to 90-100 cm2) are defined as a design basis accident for the VVER plants. Due to substantial effort in the nineties there is a comprehensive understanding of the vulnerabilities of VVER plants with respect to PRISE including safety upgrading measures. The latter include early leak detection, development of EOPs along with adequate operator training. Effective backfits have been proposed including automatic isolation of the affected steam generator, increasing pressuriser spray effectiveness, I&C logic changes allowing the operators to terminate safety injection, increased core cooling storage capacity [22]. Accident analysis approaches were presented [23-24] and they indicate that available methods are sufficient for evaluation of the safety concerns.

REFERENCES

[1] “Validation Matrix for the Assessment of Thermal-Hydraulic Codes for VVER LOCA and Transients”. OECD NEA/CSNI/R(2001)4. OECD Nuclear Energy Agency, April 2001. [2] H. PURHONEN, J. KOUHIA, H. HOLMSTRÖM, “ISP 33, OECD/NEA-CSNI International Standard Problem No. 33, PACTEL Natural Circulation Stepwise Coolant Inventory Reduction Experiment”, NEA/CSNI/R(94)24, Parts 1 and 2, December 1994. [3] “Simulation of a Loss of Coolant Accident. Results of a Standard Problem Exercise on the Simulation of a LOCA”. IAEA-TECDOC-425. Vienna, 1987., pp 1-304. [4] “Simulation of a Loss of Coolant Accident with Hydroaccumulator Injection. Results of the Second Standard Problem Exercise on the Simulation of a LOCA”. IAEA-TECDOC-477. Vienna, 1988., pp 1-288.

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[5] “Simulation of a Loss of Coolant Accident with a Leak on the Hot Collector of the Steam Generator. Results of the Third Standard Problem Exercise”. IAEA-TECDOC-586. Vienna, 1991., pp 1-317. [6] “Simulation of a Loss of Coolant Accident without High Pressure Injection but with Secondary Side Bleed and Feed. Results of the Fourth Standard Problem Exercise”. IAEA-TECDOC-848. Vienna, 1995. [7] “Separate Effects Test Matrix for Thermal-Hydraulic Code Validation. Vol. I. Phenomena Characterisation and Selection of Facilities and Tests”. NEA/CSNI/R(93)14/Part.1/Rev. OECD NEA, September 1993. [8] “CSNI Integral Test Facility Validation Matrix for the Assessment of Thermal-Hydraulic Codes for LWR LOCA and Transients”. NEA/CSNI/R(96)17. OECD NEA, July 1996. [9] M. PROTZE, “RELAP5/MOD2 Post-Test Calculation of a Loss of Feedwater Experiment at the PACTEL Test Facility”, 3rd Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, October 18-20, 1994. pp 71-88. [10] V. KORTENIEMI, E. VIRTANEN, T. HAAPALEHTO, “Analysis of the PACTEL Loss-of-Feedwater Experiments”, 3rd Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, October 18-20, 1994. pp 89-96. [11] J. BELIAEV, W. LUTHER, S. SPOLITAK, N. TRUNOV, I. TSCHEKIN, “Modelling Horizontal Steam Generator with ATHLET: Verification of Different Nodalization Schemes and Implementation of Verified Constitutive Equations”, 3rd Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, October 18-20, 1994. pp 97-106. [12] YU. BELJAEV, S. ZAITSEV, G. TARANKOV, “Verification of DINAMIKA-5 Code on Experimental Data of Water Level Behaviour in PGV-440 Under Dynamic Conditions”, 3rd Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, October 18-20, 1994. pp 118-135. [13] J. HYVARINEN, J. KOUHIA, “Experimental Verification of the Horizontal Steam Generator Boil-off Heat Transfer Degradation at Natural Circulation”, 4th Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, March 11-13, 1997. pp 1-14. [14] G. TARANKOV, N. TRUNOV, B. DRANCHENKO, W. KAMIAGIN, “Direct Measurements of Secondary Water Inventory of Steam Generator PGV-213 in Operation”, 4th Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, March 11-13, 1997. pp 15-18. [15] K. SAKATA, Y. NAKAMURA, N. NAKAMORI, T. MIZUTANI, S. UWAGAWA, I. SAITO, T. MATSUOKA, “Advanced Technologies on Steam Generators”, 4th Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, March 11-13, 1997. pp 19-25. [16] H. PURHONEN, M. PUUSTINEN, “Integral System and Horizontal Steam Generator Behavior in Noncondensable Gas Experiments with the PACTEL Facility”, 5th Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, March 20-22, 2001. pp 29-41. [17] L. PERNECZKY, A. GUBA, GY. ÉZSÖL, L. SZABADOS, “15 Years of the Hungarian Integral Type Test Facility: Horizontal SG Related PMK-2 Experiments”, 5th Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, March 20-22, 2001. pp 42-63. [18] S. LOGVINOV, V. TITOV, M. NOTAROS, I. LENKEI, “Thermal-Hydraulics of PGV-4 Water Volume During Damage of the Feedwater Collector Nozzles”, 3rd Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, October 18-20, 1994. pp 33-48. [19] A. AGEEV, R. VASILEVA, B. NIGMATULIN, V. TITOV, G. TARANKOV, “Data Bank on Hydrodynamics, Thermal Tests and Tube Temperature Regimes of PGV-4 and PGV-1000 Natural Steam Generators”, 3rd Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, October 18-20, 1994. pp 143-165. [20] O. MATAL, J. KLINGA, T. SIMO, “Primary Collector Wall Local Temperature Fluctuations in the Area of Water-Steam Phase Boundary”, 3rd Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, October 18-20, 1994. pp 136-142. [21] A. BLAGOVECHTCHENSKI, V. LEONTIEVA, A. MITRIUKHIN, “Coolant Stratification and its Thermohydrodynamic Specifity Under Natural Circulation in Horizontal Steam Generator Collectors”, 4th Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, March 11-13, 1997. pp 156-162. [22] J. MISAK, “Primary to Secondary Leaks in WWER Nuclear Power Plants”, 5th Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, March 20-22, 2001. pp 181-188.

79 Ivan Tóth (KFKI, Hungary)

[23] P. MATEJOVIC, L. VRANKA, M. BACHRATY, “Support Analysis for Emergency Operating Procedures of VVER-440 V213 Units”, 5th Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, March 20- 22, 2001. pp 199-211. [24] O. POTRAVKA, “Analysis of Emergency Operating Procedures Effectiveness for Core Damage Prevention Using Computer Code RELAP for Nuclear Power Plants with VVER-1000/B-320 in Reference to Primary to Secondary Circuit Leak”, 5th Int. Seminar on Horizontal Steam Generators, Lappeenranta, Finland, March 20- 22, 2001. pp 212-226. [25] H. KARWAT AND H.E. ROSINGER, “The Status of the Bubbler-condenser Containment System for the Reactors of the VVER-440/213 Type”, OECD NEA Report No. NEA/CSNI/R(98)13, February 1998. [26] WOLFF, H.; SWEET, D., “EU TSO Support to CEEC and CIS Nuclear Regulatory Authorities and their TSOs in the Safety Related Evaluation of the VVER-440/213 Bubbler-condenser Experimental Qualification Project” Final Report, Volumes 1 and 2 RISKAUDIT Report N° 228, December 1998. [27] KARWAT, H.; WOLFF, H., “Bubbler Condenser Related Research Work - Present Situation” NEA/CSNI/R(2001)3, September 2000. [28] “Answers to Remaining Questions on Bubbler Condenser”. Activity Report of the OECD NEA Bubbler- Condenser Steering Group, NEA/CSNI/R(2003)12, January 2003. [29] “OECD-IAEA Paks Fuel Project”, Final Report (to be published as a joint OECD-IAEA report [30] “VVER-1000 Coolant Transient benchmark (V1000-CT)- Volume I : Main Coolant Pump (MCP) Switching On - Final Specifications”, NEA/NSC/DOC(2002)6 [31] “VVER-1000 Coolant Transient Benchmark Phase 1 (V1000CT-1) Vol. 2: Summary Results of Exercise 1 Point Kinetics Plant Simulation”, NEA/NSC/DOC(2006)5 [32] “VVER-1000 Coolant Transient Benchmark Phase I (V1000CT-1), Volume 3 : Summary Results of Exercise 2 on Coupled 3-D Kinetics/Core Thermal-hydraulics”, NEA/NSC/DOC(2007)18

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