21st Symposium of AER on VVER Reactor Physics and Reactor Safety Dresden, September 19 - 23,20ll

Gonference organizer:

Institute of Safety Researchwithin the Helmholtz- Zentrum Dresden- Rossendorf, Germany

VUJE, Inc., Okruzna 5, SK 918 64 Trnova, Slovakia

KFKI Atomic Energt Research Institute, Reactor Analysis Department, Budapest, Hungary

Corporate organ izations :

Budapest University of Technologt and Economics, Hungary

KFKI Atomic Energt Research Institute, Hungary

Pal

GRS, Germany

SXOOI JS a.s., Czech Republic

Nuclear Research Institute frez, Czech Republic

CEZ, a.s., Dukovany NPP, Czech Republic

Slovak University of Technologt in Bratislava, Slovakia

NRC "Kurchatov Institute", Russia

SNIIP Atom Ltd. Moscow, Russia

SSTC NRS Kiev, Ulvaine

Kozloduy NPP, Bulgaria

INRNE Sofia, Bulgaria

Ben-Gurion University of the Negev, Israel

Oregon State University, USA

Fortum Power ond Heat Ltd., Finland 21st Symposium of AER on rlZ'JFI- ---- WER Reactor Physics and Reactor sarety- Dresden, September 19 - 23,20ll 6 ^""o"'I ZENTRUM DRESDEN R0SSENDoRF I Papers and Presentations

Introduction 0.1 Opening and Welcome, S. Kliem, IIZDR, Germany 0.2 Overview on the Institute of Safety Research within the Helmholtz- Zentrum Dresden- Rossendorf, S. Kliem, HZDR, Germany 0.3 Publication possibilities for AER scientific papers, I-P"p*ll Pr.*tl A. Aszodi, Budapest University of Technologt and Economics, Hungarj- Session I - Sssion chair: T. Simeonov 0.4 Safety assessment of the German NPPs after the Fukushima accident, l-Prp*ll Pr*"rtl A. Pautz, GRS, Germany Topic 2 Reoctor physies ryefirnen$ arrd code vultdatton

2.1 "Full-Core" - VVER-440 core periphery power distribution benchmark proposal, TP"p*ltP.esentl V. Kr!,sl, P. Mikold|, D. Sprinzl, J. Svarny, SXOnd JS a.s., Czech Republic 2.2 Calculations of fission rate distribution in the core of VVER-1000 mock-up on the LR-0 reactor using altemative methods and l- PafillF.effil comparison with results of measurements, S. Zaritskiy, A. Kovalishin, T. Tsvetkov, NRC "Kurchatov Institute", Russia 2.3 Qualification of the APOLLO2 lattice physics code of the NURISP platform for VVER hexagonal lattices, i P"pdE"s"dl A. Keresztfiri, G. Heg,ti, A. T6to, KFKI Atomic Energt Research Institute, Hungary Session 2 - Session chair: A. Kereszt{ri 2.4 The simplified P3 approach on a trigonal geometry in the nodal reactor code DYN3D, fPry"f[p'"'dl S. Duerigen, E. Fridman, I{ZDR, Germany 2.5 Solution of the "MIDICORE'WER-1000 core periphery power distribution Benchmark by KARATE and MCNP, P"fillne'*il E. Temesvdri, G.Hord6sy, G. Hegti, Cs.Mardczy, KFKI Atomic t Energt Research Institute, Hungary 2.6 Solutions for the task I and task 2 of the benchmark for core burnup calculations for a WER-1000 reactor, P"p*ltP.*"'tl T. Ldtsch, V. Khalimonchuk, A. Kuchin, I fW SUD Industrie Sewice GmbH, Germany

21st Symposium of AER on WER Reactor Physics and Reactor Safety Dresden, September l9 - 23,2011 2.7 of a new 3_D neutronics model in ApRos, ?";;::il"f#,XX#;" Tp"pn[p,*."t1 2.8 SP3 solution versus diffusion solution in nodal codes - which improvement can be expected, B. Merk and S. Duerigen Session 3 - Session Chair: I. Ovdiienko Tapic I Spectral ond corc calealations l.l Information about AER WG A on improvement, extension and validation of parametrized few-group libraries for VVER 440 and l-P"p* ll P'"'*tl VVER IOOO, P. MikoldJ, SKODA JS a.s., Czech Republic 1.2 FIELIOS: Application for criticality limits assessment, T. Simeonov, Studsvik Scandpower GmbH, Germarry tTrffllP'".*tl 1.3 A comparison of the FA's models with the detailed and simplified description of the design elements in calculations by MCU code, t-tr"-p"Il-P*'il] A.S. Bikeev, S.I/. Marin, E.A. Sukhino-Khomenko, NRC "Kurchatov Institute" Moscow, Russia 1.4 Development of multi-group spectral code TVS-M, A.P. Lazarenko, A.V. Pryanichnil

7.I Information about Group H meeting, L.K Shishfuiv, NRC "Kurchatov Institute" Moscow, Russia l-T"p*llffil

2lst Symposium of AER on WERReactorPhysics and Reactor Safety Dresden, September 19 -23,2011 7.2 Estimation procedure for engineering margin factors of VVER fuel cycles, t P"dll P'*"'tl T.G. Dementiev, D.A. Oleksuk, L.K. Shishknv, NRC "Kurcholov Institute " Moscow, Russia Session 6 - Session Cbair: R 7.ejac Topie 3 Corc design, manitor@ undfael managemcn, 3.1 AER Working Group C activities in 2011 (oral presentation, only), P"p.IlP'.t*tl I. Nemes,Paks NPP, Hungory I 3.2 In-core control system modernization experience on WWER-I000 units, canceled A. Bylmv, SNIIP Atom Ltd. Moscow, Russia 3.3 Status and prospects of the core surveillance system SCORPIO-VVER--fP"fill in Czech Republic and Slovakia, - P."*ttl J. Molnar, R. Vocka, Nuclear Research Institute frez, Czech Republic 3.4 Role of the team of scientific and technical commissioning support during Mochovce NPP Unit3&4 commissioning, I P"pe'llE J. Hermanshi, M. Prachdr, M. Sedldcek, VUJE Inc., Slovakia "'tl Seesion 7 - Session Chair: J. Molnar 3.5 AER Working Group B activities in20ll, t-F-rfillF..*"tl P. Darileh VUJE Inc., Slovakia 3.6 i5m-cycle option forNPP Paks operation, I. Nemes, Pal6 NPP, Hungary TPadll P*'""t] 3.7 Experience of TVSA fuel implementation at Kozloduy NPP, P."r*tl K. Kamenov, AL Ksmenov, D. Hristov, Kozlo&ty NPP, Bulgaria fP"p"'-ll 3.8 Implementation of 5-year fuel cycle strategy for control fuel assemblies at Dukovany NPP, tT"p*ll P'"'*tl J.Bajgl, CEZ, a.s., Dukrany NPP, Czech Republic Session 8 - Session Chair: I. Nemes Toplc i Core design, moniturtng andfuel menagement

3.9 New practice for the evaluation of rod efficiency measurement by rod drop at the NPP Paks, t-PapdlfP.esEl I. P6s, T. Parkn, Pal

2lst Symposium of AER on WER Reactor Physics and Reactor Safery Dresden, September 19 - 23,2011 Topic 5 Spentfuel disposal, aetinide transmutalion

5.1 AER Working Group E activities in2A|l, V. Chrapciak, YUJE Inc., Slovakia fP"fill P*'*tl Session 9 - Session Chair: J. Bajgl 5.2 Criticality safety analysis of fresh and spent fuel storage and handling forNPP Mochovce using MCNP5, P..'El G. Farkns, J. Haiiik, J. Liiley, B. Vrban, M. Petriska, V. Slugert, P. I-P"p"Il Urban, Slovak University of Technologt in Bratislova, Slovakia 5.3 WER-440 with viable direction to sustainability, P. Darilek, C. Strmensky, R. Zajac, J. Majercik, VUJE Inc., Slovakia fP"perltP'.*"tl 5.4 Comparison of square and hexagonal fuel lattices for high conversion PWRs, t-P-p".ltP*ffil D. Kotlyar, E. Shwageraus, Ben-Gurion University of the Negev, Israer--- 5.5 ALLEGRO - Introduction to GFR, P. Darilek, R. kjac, VUJE Inc., Slovakia I-P"p".lt P'"*"tl Topic 7 Engineeringfactors

7.3 Account for uncertainties of control measurements in estimations of design margin factors, I P"p* ll P'"*"tl Y.G. Dementiev, V.D. Sidorenla, L.K. Shishkav, NRC "Kurchatov 'Institute " Moscow, Rus sia 7.4 Specific features of accounting for probable power excursions at periphery FAs, which are caused by FA gaps behaviour in the course ot_ VVER-1000 operation, E.F. Michailov, L.K Shishkov, NRC- "Kurchatov Institute" Moscow, Russia Session 10 - Sssion Chair: U. Rohde Toplc I Spectral end cote cslculation

1.1 I Steps ahead in the few-group cross-section library generation at the pin level, f P"p* ll P*'*tl N. Petrov, J-J. Herrero, INKNE Sofia, Bulgaria l.l2 Advanced calculation schemes and cross-section libraryr egeneration in ,- hexagonal geometry with APOLLO2, t P"p* ll P."tAl G. Todorova, N. Petrov, N. Zheleva, N.P. Kolev, F. Damian, INRNE Sofia, Bulgaria Topic 4 Reaetor dynuntics and safety anatysis

4.1 AER Working Group D on WER safety analysis - report of the 2011 meeting, S. Kliem, HZDR, Germany l-P"p*ll P'*."t] 4.2 Preliminary results of the seventh three-dimensional AER dynamic benchmark problem calculation. Solution of problem with DyN3D- and RELAp5-3D-codes, f"@lT.".ertl ii x . I M. Bencih J. Hddeh Nuclear Research Institute Rez, Czech Republic

2lst Symposium of AER on WER Reactor Physics and Reactor Safety Dresden, September 19 - 23,2011 Session 11 - Session Chair: A. Asz6di 4.3 Simulation of AER-7 benchmark with the coupled code DYN3D/ATHLET, lP"filin"ffi4 Y. Kozmenknv, S. Kliem, HZDR, Germany 4.4 COBAYA3ELICA4 vs. DYN3D/FLOCAL Solutions ofthe VVER- 1000 MSLB benchmark, I Prp*lln"seE I. Spasov, T. Tzqnov, N. Kolev, J. Hadeh INRNE SoJia, Bulgaria 4.5 Modeling the spatial distribution ofthe parameters of the coolant in the reactor volume, l-P"p"'ll P'.r*tl S.P. Nilwnov, NRC "Kurchotov Institute" Moscow, Russia 4.6 Comparisons with measured data of the simulated local core parameters by the coupled code ATHLET-BIPR-WER applying a new enhanced model of the reactor pressure vessel, tP"e.Il P'"*'?l S.P. Nikonov,I. Pasichnyk, K.Velkov, GRS, Germany

Session 12 - Session Chair: S. Kliem 4.7 WWER safety analysis in case of simultaneous positive reactivity introduction by control rods withdrawal and pure condensate injection, h@l pr.r*tl A. Kuchin,I. Ovdiienka, V. Khalimonchuk, S,SZCNRS Kiev, (Jlraine L- 4.8 Study on severe accidents and countermeasures for WER-1000 reactors using the integral code ASTEC, TP"p"Itp'.'.E P. Tusheva, N. Reinke, F. Schcifer, E. Altstadt, S. Kliem, HZDR, Germany 4.9 Computational and design considerations for innovative Light Water Reactor with natural circulation, I P"p". llp."*,nil A. Soldatov, T. S. Palmer, Oregon State University, USA 4.10 Studies on boiling water reactor design with reduced moderation and analysis of reactivity accidents using the code DYN3D-MG, U. Rohde, HZDR, Germany fP"p*llffil Session 13 - Session Chair: T. Lahtinen 4.11 Application of DYN3D/ATHLET and DYN3D/RELAP5 coupled codes to simulation of RUTA-70 reactor with CERMET fuel, TF-"fillT'-"'*tl Y. Kozmenlav, Y. Baranaev, A. Glebov, U. Rohde, mDR, Germony Tople 6 Nuclear applieations of thrce dtmensional themul hydraallcs 6.1 AER Working Group G activities in 2011, TP"dltF'**tl A. Asz6di, Budapest University of Technologt and Economics, Hungary 6.2 Investigation of coolant mixing in WER-440/213 RPV with improved t;*;i::;X|;#,' Budop",t university of rechnotosr and Econom,"r @|Ftetetttl Hungary 6.3 New studies of the natural convection around the fuel rod model of the BME training reactor with PIV/LIF technique, Tp-*.ll-p-**tl- R. Szijdrt6 , A. Aszddi, B. Yamaji, Budapest University of Technologt | -r-- rlJl ond Economics, Hungary

2 1 st Symposium of AER on WER Reactor Physics and Reactor Safety Dresden, September 19 - 23 , 201 I Session 14 - Session Chair: IC Yelkov Tapic 6 Nuclesr appllcattons of three dimensional thermal hydraulies

6.4 Effect of spacer grid mixing vanes on coolant outlet temperature distribution, t-P"perll P'esentl T. Rrimci, T. Lahtinen, Fortum Power and Heat Ltd., Finland 6.5 Analysis of coolant flow in central tube of WER-440 fuel assemblies, _ G. Zsiros, S. T6th, A. Asz6di, Budapest []niversity of Technologt l-l@l|E*dl and Economics, Hungary 6.6 Analysis of Cl experiment on UPTF facility with an advanced ATHLET model, [P"fillF"ffil I. Pasichnyh K. Velkov, S.P. Niknnoy, G.i?,S, Germany 6.7 A proposal to continuation of VVER-440 fuel assembly head benchmark, tP"p*llP**"tl A. Asz6di, S. T6th, Budapest University of Technologt and Economics, Hungary Session 15 - Session Chair: IC Velkov Tapie 5 Spentfuel disposal, ac"tinide transmutation

5.6 Summary of 13th session of the AER Working Group F- "Spent fuel transmutations" and 4th meeting of INPRO project RMI - "Meeting l-P"p"IlP."ffil energy needs in the period of raw materials insufficiency during the 21st cenfury", V Leleh Nuclear Research Institute Rez, Czech Republic 5.7 Strategies of the future technological development, V. Lelek, Nuclear Research Institute Rez, Czech Republic lP,p.'llffi4

2l st Symposium of AER on WER Reactor Physics and Reactor Safety Dresden, September 19 - 23,2011 2. 1 INFORMATION ABOUT AER WG A ON MPROVEMENT, EXTENSION AND VALIDATION OF PARAMETRZED FEW-GROUP LIBRARIES FOR WER 440 AND WER 1OOO

Pavel Mikol65 SfOOn JS a.s., Orlk266,316 06 Plzei, CzechRepublic Tel: +420 j78042828, Fax: +420 378042305, E-mail: [email protected]

ABSTRACT

Joint AER Working Group A on ,,Improvement, extension and validation of parameteized few-group libraries for WER-440 and WER-1000* and AER Working Group B on ,,Core design" twentieth meeting was hosted Uy SKOOR JS a.s. in Plzefi (Czech Republic) during the period of 9th to 1Ift May 2011.

tn totat 19 participants from 7 mernber organizations were present and I I papers were presented.

Objectives of the meeting of WG A are: Issues connected with spectral calculations and few-groups libraries preparation, their accuracy and validation.

Presentations were devoted to some aspects of transport and diffrrsion calculations and also benchmarks dealing with WER-1000 and WER440 core periphery power tilt.

Zs. Szicsdnyi (co-author I. Nemes) presented "Investigation of highly enriched fuel for WER-440 application using different geometry and enrichment profile", R Zajac (co- authors P. Daiilek, C. Strmensb!,and J. Majerilk) spoke about 'VVER-440 with IMF Safety Analysis of Control Rod Ejectior", G. Horddsy (co-authors Gy. HegtL Cs. Mardczy and E. *MIDICORE" Temesvdri) presented paper called "Preliminary Solution of the WER-1000 Core Periphery Power Distribution Benchmark by I(ARATE and MCNP", P. Daiilek (co- authors M. Antal and D. Buniek) spoke about "First utilization of GD2 fuel with enrichment 4,87yo at SE-EBO", P. Mikoldi (co-outhor P. Veselj,) presented paper called "WER-440 Fuel Cycles with lncreased Enrichment and Uranium Mass" and D. Sprinzl (co-authors V. Krysl, P. Mikolde and J. Svarn!,) provided a "Full-Core" WER-440 pin power distribution calculation benchmark proposal".

Future activities are also shortly described in the end of the paper. INTRODUCTION

Joint AER Working Group A on ,,Improvement, extension and validation of panmeteized few-group libraries for WER-440 and WER-1000" and AER Working Group B on,,Core design" nineteenth meeting took place in Plzeri (Czech Republic) during the period of 9ft to I ln May 201l.

In total 19 participants from 7 member organizations were present (List of participants is attached in the end of this paper) and 11 papers were presented (List of presentations is also attached).

Objectives of the meeting of WG A are: Issues connected with spectral calculations and few-groups libraries preparation, their accuracy and validation.

PRBSENTATIONS

Six papers were presented in frame of WG A:

l) Mn Zs. Szdcsenyi (co-author I. Nemes) introducted "Investigation of highly enriched fuel for WER-440 application using different geometry and enrichment profile".

First, he explained the goal of investigation (to reduce the fuel cost (five year cycle), to increase the length of cycles), showed the frame parameters of investigation (longer cycle (15-18 month)o /higher infinite multiplication factor (higher enrichment), higher burnable absorber content (6 pin with Gd)/ with upgraded reactor power (1485 MW) takrng into account reactor physical parameters limits (assembly power, pin power, linear heat rate and sub-channel outlet temperature).

Different pin (and pellet) designs were investigated in several types of fuel assemblies. The most promising is a FA marked as G2.5 (gd-2.5) with the following features: Cladding outer diameter (tvel) 8.9 mm (tveg 9.1 mm), no central hole (weg 1.2 mm) with 2"U radial enrichment profilation (4.95 w%'3tU,4.6 wYo23sU,4.4 wYo and 4.2 wYo235U with GdzO: for reducing inner pin power non-uniformity, see Fig. 1 and Fig. 2.

Then, different core loading strategies were investigated, namely: Equilibrium cycle calculation I 12 month / 72 assemblies / g-2.5 (gd-2n) Equilibrium cycle calculation tr 15 month / 90 assemblies / g-2.5 Equilibrium cycle calculation III 18 month / 114 assemblies / g-2.5 Equilibrium cycle calculation IV 15 month / 96 assembly I gd2 m Dffirent characteristics offuel cycles were shown. (Boric acid concentration, sub-channel output temperature, max. linear power, pin power,...)

Fig. I

Maximal normalized pin power

1,14

1,12

1 ,'t

1,08

1,06

1,U

1,O2 0 5000 10000 15000 20000 25000 30000 350t Kiigis (MWnap/tU) Fig.2

Also,fuel cost and mmimal burn-up evaluation were included into presentation. Conclusion:

The most promising is the type gd-2.5. Possess the advantage of KARKAZ without its disadvantage Iligh infinite multiplication factor Low power and temperature peaks Good behaviour under the cycle Low fuel cost at the 12-15 month cycle

2) Mr. R. Zajac (co-authors P Daiilek,C. Strmensh!,andJ. Majeriik) presentedastudycalled "WER-440 with IMF Safety Analysis of Control Rod Ejection"

First, he showed a schema for WER-440 fuel cycle with CFA, then the loading pattern with CFAfuel assembly, includingfuel loading assortment with CFA.

Full core evaluation coveredz - Critical boric acid concentration and goup 6 position during the cycle - Reactivity coefficients during the cycle - Effective fraction of delayed neutrons - Average neutron lifetime - Efficiency of ejected control rod (peripheral group No6 rod, N:0olo, TM:260'C)

From fuel cycle parameters and indicators the most important is saving of Pu amount to be sent in the repository, namely: 4.87 w%o235u+3.35 w% Gd2o3* CFA Pu disposed in the repository [kg/Twtr"l 29.7219 9.4769

Control rod ejection: Four cases were evaluated (BOC or EOR for zero power or full power at different initial group 6 position).

For Case A (BOC, power 0.001 MW, CFA6 position 50 cm, core input temperature 260 "C) an additional information about neuton power, fuel temperature and cladding temperature during selected time interval was shown.

Also acceptance criteria (enthalpy, temperature) summary for U-pins and Mo-pins were provided.

A new fuel assembly design (advanced IMF WER-440 assembly (EMOC) has been proposed, see Fig. 3. C 4,95 o/o [T-2JS (72) o 4.6 o/o tt-2JS (12) C 4.4 o/o [1-235 (6) c 10.5 % Pu in l\Io + 3.0 o/o cdpr {6) C 10 % Pu in ilIo (12) c 14 % Pu in ]\to (18) Fig.3

Some comparisons of advanced Mo WER-440 assembly with uranium WER-440 235U+3.35 assembly (4.87 wo/o wo/o GdzOt) were also provided (k-inf course, maximum relative pin power during burn-up, relative pin power maps of the advanced WER-440 Assembly (EMOC) (BOL, burn-up 50000 MWd/tU).

Conclusion:

CFA (5% Pu): Equilibrium cycle with CFA's too short No problems at full core evaluation (stationary states) Two criteria were not fulfilled at one of four analyzed cases of control rod ejection

Directions of problems solution: Higher Pu content -+ CFAI (10-l4yo Pu)o more "steady" pin power distribution, lower in-assembly power peaking --+ CFAI Energetic potential of the new EMOC assembly is comparable and close to uranium 23sU+3.35 WER-440 assembly (4.87 woh w%o GdzOr) Reduction of ejected rod efficiency by core loading optimization Specification of conservative ejected rod efficiency for cycles with CFAI's Specification of conservative gas gap heat transfer coefficient

3) Mr. G. Horddsy (co-authors Gy. Hegti, Cs. Mardczy and E. Temendri) presented paper (MIDICORE" "Preliminary Solution of the - WER-1000 Core Periphery Power Distribution Benchmark by KARATE and MCNP" based on: ,,'MIDICORE'wER-1000 CORE PERIPHERY POWER DISTRIBUTION BENCHMARK PROPOSAL" introduced by D. Sprinzl on 20th Symposium of AER on WER Reactor Physics and Reactor Safety in Finland 2010. Objective of this worlc to give reference for validation of accuracy of pin-by-pin power distribution at WER-1000 core periphery (see Fig. 4).

Cailogram of 'midicore'

E cote n 'midicore- T water A n steel mixture L-*, il water+steel

Fig.4

Together four different FA types (4200, E4086, P36E9 and P40E9 with different enrichment, profilation and number of tvegs) were put in this small core.

2D benchmarkfor cold state geometry, T:600K.

Material composition is homogeneous in axial direction, except the steel-water mixture at a small part of the reflector. Here a) A reference model with periodic steeVwater structure 20138 mm b) Standard model with smearing over volume

To be calculated:

- ICr - Pin by pin power distribution for assemblies 6, 7 and 9 - Integral fission power in assemblies l-10

Preliminary calculation at KFKI-AKI by MCNP and KARATE code systems:

MCNP: version MCNP5 1.40 Libraries: ENDF/B-VI for most isotopes, .14c for uranium isotopes, .62c and.66c for the others, but ENDF/B-V for Fe, Co and Ni (.50c and .55c). KARATE: based on ENDF/B-VI.

MCNP results: IQn SKODA: 1.04538, std.:0.00003 AEKI: 1.04411, std.:0.0001

Differences between the assemblies integrated power distributions No. l-10: 0.I% - 0.46 yo, average: 0.3 %. Difference of pin power distributions: < 1 % (Differences in individual pins for three FAs investigated were also provided.)

Preliminary Resultsfrom KARATE (1) MIDICORE WER-1000 Benchmark FA K;6comparison (see Table l)

Name of Kior calculated St.Dev. Kiol calculated by FA bv MCNP MCNP MULTICELL A200 1.07861 3 .98-04 L .07 L67 A4OE6 L.2228L 4.08-04 t.2L420 P3689 1.16005 4.08-04 1.15152 A4OE9 1.19049 4.28-04 1.18051

Table l.

Preliminary Resultsfrom KAMTE (2) MCNP, reference model, hn:l.04520 KARATE(GLOBUS) model k"n :1 .g3rr4, diference :1.5 -6 Yo.

Conclusion

Preliminary calculations show good agreement between Monte Carlo calculations, but not so good between Monte Carlo calculation and production code.

4) Mr. P. Daiilek (co-authors M. Antal and D. Buniek) presented paper 'sFirst Utilization of GD2 Fuet with Enrichment 4.87 w%o '3sU at SE-EBO".

A short history of EBO units was given in the beginning of his presentation. Loadedfuel:

- Up to 2006 profiled fuel with mean enrichment 3.82 wYo23su - Since 2006 profiled Gd-2 fuel with mean enrichment 4.25 wYo235U - WA and 3.84 wyo235tJ CA - Planned since 2012: profiled Gd-2 fuel with mean enrichment 4.87 w%o23slJ 23sU 23sU - Contract on recent fuel 4.25 wo/o and 3.84 wyo was valid up to 20 I 0 At 2008 - TVEL proposal: - Contract prolongation on already used fue|4.25 wyo23su and 3.84 wo 235U - New contract conclusion on new fuel with mean enrichment 4.87 wYo235LJ. Next, 4.87 wYot"U fuel licensing sequence was specified. FA basic characteristics were provided. Fuel pin basic characteristics were specified. Selected fuel exploitation limits were given.

Cycle choracteristics ( I /2) Reactor power I47 1.25 MW t Core cycle lengths: Unit 3, cycles 29-40 309.8 - 339.4 duy"u Unit 4, cycles 28-39 300.1 - 339.6 duy"u

Nb of necessary fresh FAs:54 - 66 WAs, 6 - 6+1* CAs, 1* - CA 1,6 woA235U l*t generation.

Maximal reached (calculated) burn-up at proposed cycles: - wA 58.0 MWd/kgU - CA 58.0 MWd/kgU - pin without Gd 67.2 MWd/kgU - pin with Gd 62.8 MWd/kgU - pellet without Gd 75.2 MWdlkgU - pellet with Gd 70.2 MWd/kgU

Cycle characteristics (2 /2) - Mean interval between fuel re-loadings 12 [month] - Mean FA cycle prolongation 5.13 - 5.18 [year] - Maximal FA cycle prolongation 6 [year] - Initial critical HrBO: conc.- l00yo, eq. pois. 5.30 - 5.55 tdtgl - Shutdown H:BOr conc. for refueling 12.01 - 12.29 lglkgl

Shift of shutdown H3BO3 conc. at all volumes: 12,0'-'+ 12,8 tdtg].

Lengths of cycles of Units i and Unit 4 were also shown. Graph offuel burn-up andfreshfuel consumptionwere also given. Main cycle parameters for unit j, cycle 29, unit 4, cycle 28, Unit 4, cycle 39 and main periphery parameters were displayed.

Conclusion

Experience with new Gd fuel, including some historical view was provided.

5) Mr. P. Mikoldi presented paper *WER-440 Fuel Cycles with Increased Enrichment and Fuel Mass".

He stated in his presentation that testing of different fuel assemblies (FA) designs and different enrichment of fuel including fuel cycles features has been ca:ried out at SfOpa fS per order of (PZ a.s in years 2006-2010.

It was tested: - Different placement of FR with Gd BA and different content of GdzOr - Enhancement of UO2 mass (bigger fuel pellet diameter with/without central hole) - Higher enrichment and different radial profilation in an FA - Another design changes (cladding thickness, FP pitch, FA shroud thickness etc., including essentially different FA design [,,Karkaz. . . "] ).

It was confirmed that: - Better intra-assembly power distribution due to radial profilation of enrichment enables better pin-to pin power distribution in the core. - This enables L3P loading pattern with longer fuel cycle, or it leads to the fuel savings.

Results: - Design of an FA with higher enrichment. Together, it is possible to think about changes in fuel rod geometry or/and fuel pellet geometry.

Some otherfindings: - Temporary content of GdzOr in FP with Gd BA is suflicient for one-year cycles (maybe too hish) - Thickness of fuel assembly shroud must be preserved; also pin pitch must be preserved (technical reasons) - Radially uniform effichment in a FA would be suitable from the burn-up point of view, but from the radial power distribution point of view is not possible (must be lower at FA shroud) - Fuel mass enhancement or its enrichment is possible, but it is an economical and licensing question.

Dffirent FA types were tested.

In the end, the following design was selected: FA ,rQFS6'with an average enrichment 4.757143wVo23sU (see Fig. 5).

o 4.95o/o Ut35(84) o 4.4"/" U235(30)

@ 4.4o/o 1J235 + 3.35o/o Gd2O3(6)

C 4.2o/o u235(61 o centr5lnitrubka

Fig.5 Karvalues and pin power non-uniformity for different FAs were also shown.

Fuel cycles 2lh - 34! of modified project for 3'd Unit of Dukovany NPP were compared. 23tU, (Gd-2M FA 4.38w% CA 4.25w%o235U, existing FP geometry.) - Different FAs were loaded in these project loadings without any changes - Prolongation of fuel cycles (FPD) due to loading of FAs of Gd-2 type higher enrichment and higher fuel mass was observed.

The other characteristics were also evaluated: - Fuel cycles - Boric acid course - cycles with different FAs and double power - Fuel cycles - Frn (I(o) - cycles with different FAs and double power - Fuel cycles - Fas (K) - cycles with different FAs and double power - Fuel cycles - Fq - cycles with different FAs and double power

Optimized fuel cycle (34) - FAS age and course of Fon - cycle with,,QFS" FAs Optimized fuel cycle (34) - FAs age and course of Fa1 - cycle with ,,SLF" FAs (average enrichment 4.87 wYo235U - for SE)

Licensing of ,,new" FAs was shortly commented.

Changes in enrichment, pellet geometry and FR are possible from the N-PH/TH viewpoint, but they are connected with the serious licensing problems: - Thermo-mechanical behavior evidence of a new RF design, theoretically, experimentally and field proved (full pellet, thin cladding etc.) - Next enhancement of fuel bum-up; Regulatory Body expects experimental results (the same as for WER-1000 tuel).

Problems with fuel storing with high enrichment and burn-up. Existing criteria are limiting, acceptance of criteria like,,Partial Boron Credit", ,,Burn-up Credit". Temporary fuel is on limits. (Spent fuel pool, Container CASTOR etc.) - Expectation connected with successful operation of Dukovany NPP on up-rated power. - Changes in access of Regulatory Body at conditions of critical view on nuclear power industry.

Conclusion

- FA of ,'QFS" type with higher enrichment enables from the viewpoint of N-PH/TH an effective 5-6-years cycle at nominal power of 1444 MWtn, eventually at power 1471 (1485) MWg, at expected fuel cycle lengths.

- FA ,'QFS" has lower average fuel enrichment then FA ,,SLO". (nZ would prefer it, because due to better characteristics of power distribution it enables more optimal type of fuel loading strategies.

- Acceptance of FA TTQFS" requires: - An economical evaluation - Resolving questions connected with licensing process. 6) Mr. D. Sprinzl (co-authors V. Krysl, P. Mikoldi and J. Svarnf) presented paper called o"Full-Core' WER-440 pin power distribution calculation benchmark proposal".

He explained why to deal with a ,,Full-Core" benchmark for WER-440, then benchmark specification (general, fuel assemblies (FA's), geometry and suggestions) was given.

Generally:

- 2D-calculation of selected AZ state with fresh fuel with 30" reflection symmetry, see Fig. 6 - Reference: Monte Carlo calculation (probably MCNP) - Solution: any (own) macro-code calculation.

i;i.\ita rI; ::,':-.a:a:: 5'::ii:lr

Fig.6

Reference calculotion' Monte Carlo code.

2D calculation: Cold state geometry and densities of materials (except for coolant) Coolant density for 558.15 K and 12.3 MPa Temperature of all materials 558.15 K Loading pattern was selected.

Next were provided: - Fuel assemblies specification - Fuel pins specification - Materials' specification - Geometry - Suggestions - Summary Conclusion

A benchmark for calculation of power distribution in WER-440 coreo especially near the radial reflector was specified.

CONCLUSION

Six paperc were prcsented on lryG A, which are devoted to different topics.

Effort was concentrated onfuel assembly design

Proper determinafion of boundary conditions is still an important issue.

A benchmark was focused to this poinl Conclusions from the solation of this benchmark will result to improvement, if necessary, in industrial codes.

List ofNomenclature

AER Atomic Energy Research WG Working Group WER-440 Standard abbreviation of Russian PWR with nominal power 440 MW" WER-1000 Standard abbreviation of Russian PWR with nominal power 1000 MW" IMF Inner Matrix Fuel SE-EBO Slovak Power Stations - Bohunice NPP Tvel Russian abbreviation for fuel pin (without burnable absorber) Tveg Russian abbreviation for fuel pin (with burnable absorber) KARKAZ Expression for specific Russian design of a fuel assembly CFA Combined Fuel Assembly CFA6 Control Fuel Assembly /sixth group/ CA Control Assembly BOC Begrn of cycle EOR End of cycle EMOC Equilibrium Molybdenum Cycle CFAI Modification of CFA IGm, hr Effective multiplication coefficient k-inf Infinite multiplication coefficient duy"t FPD - Full Power Day FP Fuel Pin BA Burnable Absorber FA Fuel Assembly WA Working Assembly Fna (&) Maximum relative power in fuel assembly Fau (KJ Maximum relative power in fuel pin {o Maximum relative power in axial mesh in fuel pin CEZ Czechproducer and supplier of (electric) energy 2D Two-dimensional N-PH Neutron-Physical TH Thermohydraulic

APPENDIX

List of Presentations

WGA

t I. Nemes and Zs. Szdcsdryti: Investigation of highly enriched fuel for WER- 440 application using different geometry and enrichment profile t p. Daiilek, C. Strmenslry, R. Zajac and J. Majerifk: WER-440 with IMF Safety Analysis of Control Rod Ejection . G. Hordds.v, Gy. Hegti, Cs. Mardczy and E. Temesvdri: Preliminary Solution of the "MIDICORE" WER-1000 Core Periphery Power Distribution Benchmark by I(ARATE and MCNP t l,[. Antal, D. Buniek and P. Dalilek: First utilisation of GD2 fuel with enrichment 4,87oh at SE-EBO . P. Mikold| and P. Veselit: WER-440 Fuel Cycles with Increased Enrichment and Fuel Mass t l/. Kr!'sl, P. Mikold|, D. Sprinzl and J. Svarn!': "Full-Core" WER-440 pin power distribution calculation benchmark proposal

WG-B

. M. Fiola: Calculation of Selected Unit Start-up . M. So1ek. J. Svarnlt and V Krj,s/; Automated Computation of Validation Test Case for MOBY-DICK Code . J. Svarni,; Unification of the Engineering Factors Methodology Component of WER-1000 and WER-440. . K. Katovsk!, and J. Prehradn!,: WER-440 Loading Patterns Optimization Using ATHENA Code . J. Bajsl; Five Years Fuel Cycle Strategy for CFAs at Dukovany NPP List of Participants

Name Company Mr. Botond Beliczai Paks NPP

Mrs. Csilla Sz6cs6nyi Paks NPP

Mr. Zsolt Sz6as6nyi Paks NPP Ms. Emese TemesvSd MTA KFK|Atomic Energy Research lnstitute Mr. G6bor Hordosy KFKI Atomic Energy Research lnstitute Mr. S6ndor Patai.Szab6 TS Enercon Ltd Mr. Radoslav Zajac WJE, a.s.. Mr. Petr Daiilek VUJE, a.s. Mr. Josef Bajgl dEZ, a.s., Dukovany NPP Mr. Martin Fiala dU, a.s., Dukovany NPP Ms. Julie Krskov6 6U, a.s., Dukovany NPP Ms. Monika Mervartov6 GEZ, a.s., Dukovany NPP Mr. KarelKatovsla.i Sxoolls a.s. Mr. PavelMikoki5 Sxool JS a.s. Mr. DanielSprinzl Sxoon JS a.s. Mr. Martin Sa5ek $xooa.ls a.s. Mr. Jiff Svarnf Sxool.ls a.s. Mr. Svatobor Stech SxooaJs a.s. Mr. PavelVeself eEZ a.s., HQ AER WORIilNG GROUP B ACTIVITIES IN 2011

Petr DARILEK VUJE Inc., Okruzna 5, SK 918 64 Trnava, Slovakia [email protected]

ABSTRACT

Review of AER Working Group B Meeting in Plzefi, CzechRepublic is given. Regular meeting of Core Design Group was organizedby SfOOa JS a.s. and held at Hotel "U Pramenri" in Plzefi, Czech Republica on May lG:-11, 2011, together with Working Group A. Presented papers (see List of papers and List of participants) covered topics as follows.

FUEL MANAGEMENT

Two papers were in agreement with original Core Design Group topic. K. Katovslcj from SKODA-JS a.s. characterised former and recent development of loading pattern optimisation code ATHENA [1]. Developed originally for WER-1000 in Temelin, code was reoriented later on NPP Dukovany with VVER-440 reactors, improved with graphical interface and tested widely. Code includes up to 6 optimisation methods and is geometry and macrocode "almosf independent. Recent development is focused on code performance improvement by parallelisation, utilisation of new up-to-date optimisation algorithm and preparation of version again for VVER-1000. J. Bajgl from NPP Dukovany described situation of fuel cycle development at Dukovany NPP with usage of specific fuel assemblies Gd-2M(+) (4,38%), Gd-2X (4,76%) and Gd-2+ (4,25yo) at 5-year fuel cycle [2]. He continued with tendency to reach full 5-year fuel cycle also for control fuel assemblies with emphasis on symmetry of core loadings. Practical example of implementation at EDU Unit 3 was shown. WER ENGINEERING FACTORS

J. Svarnf from SKODA-JS a.s. compared at the beginning engineering factors of reactors WER-440 and WER-1000 with concentration on its methodology component [3]. Later he treated three ways of VVER-440 engineering factors adaptation to WER-1000 ones. He pointed out, that mentioned theoretical approaches were tested based on experimental data from NPP Dukovany Units I and 3 (SCORPIO) and MOBY-DICK calculations. It was concluded, that it is acceptable to decrease limiting uncertainty of more loaded WER-440 fuel assemblies and that it is possible to decrease engineering factor methodological component for MOBY-DICK and eventually to increase limits.

REACTOR CALCIILATIONS

Two papers treated practical problems, connected with WER-440 units exploitation. M. Fiala from NPP Dukovany informed about problems, experienced during start-up of Unit 3 at cycle }lb 25 [4]. Power increase was limited not by rod power but by potential pellet - clad mechanical interaction, indicated by SCORPIO module PES. Based on detailed core power distribution recalculation the problematic assembly was identified. Situation was solved by its "conditioning" during 3-4 FPD on lower power. M. Sa5ek from SXODAJS a.s. presented automated validation of MOBY-DICK code [l]. He pointed out enlarged testing and more information about the code as main motivations for the automation. Testing is based on usual exploitation data from the SCORPIO system after defined preprocessing. Calculated statistical quantities - arithmetic mean, standard deviation and extremes must not exceed predefined values. Output formats and utilisation examples were shown. Successful validation test of MOBY-DICK code was concluded.

LIST OF PAPERS tU K. Katovskj, J. Prehradnj: VVER-440 Loading Patterns Optimization using ATHENA Code t2l J. Bajgl: Five Years Fuel Cycle Strategy for CFAs at Dukovany I.IPP t3l J. Svarny: Unification of the Engineering Factors Methodology Component of WER 1000 and VVER 440 t4l M. Fiala: Calculation of Selected Unit Start-up tsl M. Sa5ek, J. Svarn;f, V. Krysl: Automated Computation of Validation Test Case for MOBY-DICK Code LIST OF PARTICIPANTS

Name Company E-mail

Mr. Botond Beliczai Paks NPP [email protected]

Mrs. Csilla Szdcs6nyi Paks NPP [email protected]

Mr. Zsolt Sz6cs6nyi Paks NPP [email protected]

Ms. Emese Temesvriri MTA KFKI AERI [email protected]

Mr. G6bor Hord6sy KFKI AERI [email protected]

Mr. Sr{ndor Patai-Szab6 TS Enercon Ltd [email protected]

Mr. Radoslav Zajac VUJE, a.s.. Radoslav.Zaj ac@vuj e.sk

Mr. Petr Dafflek VUJE, a.s. [email protected]

Mr. Josef Bajgl (EZ, a.s.,Dukovany NPP [email protected]

Mr. Martin Fiala (EZ, a.s.,Dukovany NPP [email protected] Ms. Julie Krskov6 iEZ, a.s.,Dukovany NPP [email protected]

Ms. Monika Mervartov6 CEZ, a.s., DukovanyNPP monika.mervart ov [email protected]

Mr.'Karel Katovs$f Srona JS a.s. karel.katovsky@skoda-j s. cz

Mr. Pavel MikoLi"S Srooe JS a.s. pavel.mikolas@skoda-j s.cz

Mr. Daniel Sprinzl Srone JS a.s. daniel. sprinzl@skoda-j s.cz

Mr. Martin Sa5ek Srone JS a.s. martin. sasek@skoda-j s.cz

Mr. Jiff Svarnli Srooa JS a^s. j iri.svarny@skoda-j s.cz

Mr. Svatobor Stech Srooe JS a.s. svatobor.stech@skoda-j s.cz

Mr. Pavel Veselj CEZ a.s., HQ [email protected] AER Wonxrxc Gnoup C lcrrvrry m 2011

Imre Nemes Paks NPP Ltd Hungary

Working Group C had a joint meeting with Group G in Balatonfiired, Hungary ,2-3 June, 2011. At the joint meeting 28 people from I I AER member organisations of 6 countries - such as Russia, Czech Republic, Slovakia, Germany, Finland and Hungary - participated. In the 2 days of the program l0 papers were presented in the topic of working group C. The title of papers and the list of participants are attached. At the meeting the following topics were discussed:

l. New fuel and new cycle, plans and experiences

In Mochovce, Slovakia, 4.87 o/o enriched ,2"d generation fuel was introduced. The annual fresh fuel loading is supposed to be 66-72 assemblies., the maximum expected burnup of assemblies is 66-67 MWdlkgU.

Paks NPP is examining a l5m operation of units. The system of this kind of operation is given, equilibrium and transient cycle plans under preparation.

2. Reactorphysical me&surement and evaluation problems

For several years a certain asymmetry can be observed in assembly outlet temperature signals of Dukovany Unit 4. A detailed examination program has been continued. The conclusion is that asymmetry is connected with orifices cross-sections changes and also can be impacted by crud deposit. In the last year a special measurement program was treated and the problem is finally solved. In Paks a new method for the evaluation of rod-drop measurement was introduced. The out- core detector signals are simulated using kinetic calculation and pre-calculated transfer function. The agreement between simulated and measured "reactivity" is good.

A series of assembly burnup measurement has been treated at spent fuel pond of Paks NPP. The C-PORCA simulation code was extended to calculate Cs-134 and 137 isotopes as well. A transfer function to take into account shielding of different pins in the measurement geometry was calculated at Technical University Budapest. Using all this tools the agreement between calculated and measured burnup is excellent.

3. Code validation using measured data

KARATE code at KFKI AEKI was widely validated against Paks measurement and also C- PORCA code results. KIKO-3D code was inserted into full scale simulator in Paks, and the code was back-tested to KARATE code. MIDICORE benchmark was also used in testing procedure.

Moby Dick code was also extensively tested against SKORPIO data of Dukovany NPP by Skoda JS. Well developed statistical methods were used in testing procedure.

4. In-core surveillance system developments

The SCORPIO-VVER system is under upgrade for Bohunice units, Slovakia. The upgrade contains several new service for the operators. Also the planning to install SCORPIO for new hardware and operating system has started.

A paper about the changes of in-core measurement system at Mochovce NPP was presented. The changes were initiated from HW developments as well as introduction of new fuel.

All of the listed presentations - mainly in form of PPT files - are ovailable in electronic for in case of request. AER Working Group C 2011

1. Pdlenik, Petcr Mochovce First Experience with New Nuclear FuelGd-ll-4.87%

2. Bajgl, Josef Dukovany Advances in Asymmetry of FA Coola nt Outlct Temperature 3. Urban, Pcter Mochovce Ncw Short lnformation of lnCore in Mochovce 4. Temesvdri, Emese AEKI Rccent Expcriences in the V & V of thc Core Calculations 5. 5a3ek, Martin Skoda.lS Automated Computation of thc Validation Tcst Casc 6. P6s,lstvdn Paks Simulation of rod drop measurements using the C-PORCA 7 kinetics module 7. Molndr, Jozef NRI SCORPIO-VVER EBO Upgrade 3 8. MolndqJozef NRI Porting thc System to the New HW&OS 9. Park6, Tam6s Paks Simulation of Spcnt Fuel Burnup Measurement 10. Nemes,lmre Paks 1.5m cycle option for Paks NPP operation AER Working Group C and G 201I Participants Kiss Bdla BME Attila Asz6di Dr. BME Rita Szijrirt6 BME T6th Srindor BME Koubek Vlastimil DUKOVANY Josef Bajgl DUKOVANY Martin Fiala DUKOVANY Milos Mikol6s DUKOVANY Peter Urban ENEL Peter Pilenik ENEL R6bert T6th ENEL Timo Toppila FORTUM Dmitry Posysaev GIDROPRESS Oleg Kudryavtsev GIDROPRESS Thomas Hdhne HZDR Temesvriri Emese KFKI Jozef Moln6r NRI REZ Imre Nemes PAZRT Dr. P6s Istvrin PAZRT B6na G6bor PAZRT Jrivor Erika PAZRT Park6 Tam6s PAZRT Kat6Zolthn PAZRT Martin Sasek SKODA Josef Sristek SKODA Jrln Kubacka WJE Jrin Remis VUJE AER WORKING GROUP D ON WER SATETY ANALYSIS - REFORT OF THE 20ll MEETING

S. Kliem Helmholz- Znntram Dresden-Ros sendorf Institute of Safety Research P.O.B. 51 01 19, D-01314 Dresden, Germany [email protected]

ABSTRACT

The AER Working Group D on VVER reactor safefy analysis held its 20th meeting in Stockholm, Sweden, during the period 12-13 April,20ll. The meeting was hosted by the Royal Institute of Technology (KTH) and was held in conjunction with the third workshop on the OECD/i{EA Benchmark for the Kalinin-3 WER-1000 NPP and the fifth workshop on the OECD Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) for Design, Operation and Safety Analysis of LWRs. Altogether 18 participants attended the meeting of the working group D, 12 from AER member organizations and 6 guests from non-member organization. The co-ordinator of the working group, Mr. S. Kliem, served as chairman of the meeting.

The meeting started with a general information exchange about the recent activities in the participating organizations.

The given presentations and the discussions can be attributed to the following topics:

o Code validation and benchmarking including the calculation of the OECDA.IEA Benchmark for the Kalinin-3 VVER-1000 NPP and 7ft AER Dynamic Benchmark o Thermal hydraulic analyses o Safety analyses and code developments o Future activities

A list of the participants and a list of the handouts distributed at the meeting are attached to the report. The corresponding PDF files can be obtained from the chairman.

1. CODE VALIDATION AND BENCHMARKING

1.1 OECD/NEA Kalinin-3 Benchmark

GRS and Kurchatov lnstitute proposed an international benchmark for coupled code systems based on a start-up test at the NPP Kalinin-3 (VVER-1000). It was accepted by OECDA,IEA. The considered test is the switch-off of one main coolant pump from four working at full power. Very detailed measurement data are available including time-dependent neutron power data from 64 fuel assemblies at seven elevations with a time resolution of 1 s. The benchmark was launched with the following objectives: . to verifl/ the capabilities of system codes to analyze complex transients with coupled core-plant interactions and complicated fluid mixing phenomena o to fully test the 3D neutronics/thermal-hydraulic coupling o to evaluate discrepancies between predictions of the coupled codes in best-estimate transient simulations with measured data . to perform uncertainty analysis having at disposal not only the measured values but also their accuracy

The benchmark is divided into three exercises. Exercise 1 deals with point kinetics plant simulations in order to test the primary and secondary system model responses. Exercise 2 is a coupled 3-D neutronics/core thermal hydraulics response evaluation. In this exercise only the vessel with given boundary conditions at the nozzles is considered. Best-estimate coupled code plant transient modeling is the content of exercise 3. This exercise combines elements of the first two exercises in this benchmark and is an analysis of the transient in its entirety.

The third workshop on the benchmark took place just before the working group meeting.

In his presentation C. Parisi (ENEA) gave an overview on the activities on this WER-1000 benchmark problem [1]. Preliminary results using the coupled thermal-hydraulic/neutron kinetic code system RELAP5-3D/NESTLE were presented. In the development of the input data sets special attention was given to the modeling of the reactor pressure vessel. 3D components are used to describe the flow from the inlet nozzles to the core. The final calculation results should be available by the next Kalinin-3 workshop.

1.2 The OECD/DOE/CEA WER-lfi)0 Coolant Transient Benchmark

The second phase of the OECD/DOE/CEA WER-1000 Coolant Transient Benchmark (Vlffi0CT) consisting of three different exercises includes a benchmark on a main steam line break (MSLB) in a VVER-1000 reactor. This benchmark was recently finished.

G. Alechin (GP Podolsk) presented the results of exercise 2 (core boundary condition problem) obtained using the coupled code systems TRAP-KS and KORSAR/GP [3]. TRAP- KS consists of the computer codes DINAMIKA-97 (system code for transients) and TECh-M- 97 (code for analyses of leaks) and the program modules KAMAZ (3D neutron kinetic code KARTA and lD thermal hydraulic core model KANAL) and KAMERA-V2 (2D thermal hydraulic calculations in the reactor pressure vessel). KORSAR/GP was developed on the basis of the code KORSAR/VI.I which is a system code for analyses of non-stationary processes in NPPs with water-cooled and water-moderated reactors under stationary, transient and accident conditions including a point kinetics model. KORSAR/GP includes the 3D neutron kinetic code KARTA and also program module KAMERA-V2.

In the calculations for a realistic MSLB scenario no return-to-power was predicted. Contrary to that the results for a conservative scenario with two stuck rods showed a significant return- to-power in all calculations. The results of the performed calculations and comparison with results of different codes show that the considered code packages can simulate the considered cases and can be applied for calculations. 1.3 Dynamic AER Benchmarks

C. Parisi reported about the final results in the calculations of the dynamic AER benchmarks AER-DYN-0O2 using the RELAPS-3D/NESTLE code system [1]. This benchmark was defined in the frame work of the working group in order to verify the 3D neutron kinetic core models. AER-DYN-(X)2 considers an asymmetric control rod ejection in a WER-440 reactor core at low power with a simple adiabatic Doppler feedback model. In his presentation he compared the NESTLE results to those obtained with PARCS which have been presented last year. The NESTLE results show a remarkable higher power peak also in comparison to the DYN3D results obtained by [ZDR. Increasing the number of thermal hydraulic channels into the core model improves the behavior of the calculation and reduces the maximum power reached.

A new solution for this benchmark was obtained by Kurchatov Institute using the code ATHLET-BIPR-VVER [2]. The new results show a higher power peak than the older ones obtained by the BIPR8 code.

First results of calculations of 7n dynamic AER benchmark were presented. This benchmark is the continuation of the efforts to validate systematically codes for the estimation of the ffansient behavior of WER type nuclear power plants. The benchmark concems the simulation of the re-connection of an isolated circulating loop with low temperature or low boron concentration in a WER-440 plant. Each participant should use own models for the description of coolant mixing in the lower and the upper plenum. It is expected that the different mixing options have a considerable influence on the response of the reactor core. The results on the coolant mixing obtained from a detailed CFD model (e.g. ANSYS CF)Q can serve as reference solution. The definition of the benchmark has been presented at the last AER Symposium.

J. Hadek presented preliminary results for this benchmark obtained at NRI Rez using the codes DYN3D and RELAP5-3D t8l. An existing VVER-440 model for the RELAP5-3D code was adjusted to the needs of this benchmark. E. SfjAlahti (VTT) gave an overview on the development of an APROS model for the WER-440/213 reactor including a 3D coarse mesh model of the reactor pressure vessel t9l. All earlier benchmarks have been calculated at VTf using the codes SMABRE and HEXTRAN. Now for the first time the code APROS is used recently updated with a new 3D neutron kinetic model based on the IIEXTRAN code. S. Kliem (HZDR) explained that it is foreseen to calculate the 7ft benchmark with the code systems ANSYS CFX/DYN3D and DYN3D/ATHLET tl0l. The calculation using the ANSYS CFX/DYN3D coupling is performed in co-operation with the Budapest Technical University. S. Kliem reported about the efforts conducted and problems faced in the preparation and conduction of the calculation. The DYN3D/ATHLET calculation is also in preparation and will be presented at the current Symposium.

2. TIIERMAL I{YDRAULIC ANALYSES

B. Kiss (BUTE) presented an update of the CFD model of the WER-440 reactor pressure vessel (RPV) under development at BUTE [4]. This model was continuously further developed during the last years. This very detailed RPV model contains now the following main structural elements: o inlet and outlet nozzles o guide baffles of hydro-accumulators o alignment drifts o elliptical perforated plate o planar perforated plates of brake tube chamber o brake- and guide tube chamber . core described as porous body o perforation ofreactorpit o lower grid plate of protective tube unit

Different high-order turbulence models were used to check their influence on the coolant mixing inside the RPV. The SST model gave the best agreement with the results of the coolant mixing experiments performed at NPP Paks. For a further improvement of the turbulence modeling a new model available in ANSYS CFX was used. In a fist step this hybrid turbulence model was applied in the downcomer, only.

I. Pasichnyk presented the work and the results on modeling an emergency core cooling injection experiment at the l:1 German Upper Plenum Test Facility (UPTF) using the ATHLET code [5]. For the modeling the already for several years existing approach of using a very detailed modeling of the components using the ATHLET code was used here, too. Especially the following points were highlighted:

o. New mesh and thermo fluid object (TFO) generation model for ATHLET was developed o Worked out a method for constructing the nodalization scheme o Improved scheme validation o Substantial changes in the ATHLET system code - orders of magnitude acceleration of calculation

It was mentioned that further investigations are needed to study the sensitivity of the model to various changes in the initial and boundary conditions. It is also foreseen to model the turbulent exchange in a system of parallel channels for the downcomer region.

3. SAFETY ANALYSES AI\D CODE DEVELOPMENTS

A. Keresztfri gave a presentation on the application of thermal-mechanics and thermal- hydraulics codes for the hot channel analysis of RLA events [7]. The hot channel calculation is an important final stage of the safety analysis for checking the acceptance criteria fulfillment or/and for counting up the failed fuel rods, necessary for the activity release evaluation. By the

The gas gap conductance plays an essential role in case of fast transients. Both the initial and time dependent values, which are changing during the transient, are important. In case of fast RIA transients, a thermal mechanical code has to be applied for the best estimate calculation which is needed for uncertainty analysis or even only for justifying a conseryative method. In the presented work, an on-line coupling of the thermal hydraulic hot channel code TRABCO and the thermo-mechanical code FRAPTRAN was carried out in order to explore the possibilities of such a coupling. The considered test case was a control rod ejection accident. The preliminary results, which should be checked carefully, showed that the conservative approach used up to now contains considerable reserves. The computation method of the on- line coupling was found relatively simple under INTEL FORTRAN. It was mentioned that some numerical problems appeared in the coupled calculations. The causes are some uncertainties and numerical instabilities in the codes which do not influence remarkably the stand-alone calculations. They could be solved by applying an iterative calculation process.

A. Pinegin (Kurchatov Institute) gave an overview on the statistical analysis of the determination of the core reactivity during start-up experiments with single control rod insertion in a WER-1000 reactor. It is known that the reactivity values determined on the readings of the ionization chambers depend on their location. This problem is solved by the introduction of form factors (correction parameters) into the inverse point kinetics equation. By means of the SUSA method a statistical analysis of the form factors was conducted. The following sources of errors were considered:

o Errors in multiplication factor of FA based on erors in the used nuclear constants o Errors in boundary conditions o Errors in multiplication factor for nodes with control rods based on errors in the used nuclear constants o Mechanical uncertainty in multiplication factors of FAs connected with the deviation from nominal value of the fuel parameters Errors in the sensitivity coefficients of the ionization chambers current to the fission reaction velocity in different FAs a Uncertainty in the delayed neuffon parameters a Errors in the adjoint function

The SUSA procedure was applied to 150 calculations using the NOSTRA code, where all the variations were implemented. The following conclusions could be drawn: The relative methodical error in determining the reactivity, taking into account the mentioned error sources is nearly 6-8Vo. The major contributions to the errors are uncertainties in the parameters of delayed neutrons and the boundary conditions.

4. FT]TURE ACTIVITIES

The following topics are either in progress or are of potential interest in the future activities of the working group D:

o Solution of the WER-1000 steady state benchmark AER-FCM-IOI a Calculation of OECDA.{EA Benchmark for Kalinin-3 WER-1000 NPP a Clarification of the cause and significance of mesh refinement effects on solutions to control rod ejection benchmarks a Methodology for safety analyses o Safety criteria for high burnup fuel o Uncertainty and sensitivity analysis for safety analyses a Hot pin and hot channel approximations in safety analyses a Transient fuel behaviour models and approximations for use with 3D core models o Representation of reflectors, including wide range data o Wide range representation of two-group cross section data . Application of two-group neutron kinetics data o Application of 3-D thermal-hydraulic calculations for coolant flow and mixing in the reactor vessel o Utilization of data from physical start-up experiments

It was tentatively agreed to hold the next meeting of working group D in April, 2012 in Karlsruhe, Germany, in connection with the next workshops on the OECD benchmark for uncertainty analysis in best-estimate modeling (UAM) and on the Kalinin-3 benchmark.

LIST OF PARTICIPANTS

From AER member organizations 0l A. Keresztriri KFKI Atomic Energy Research Institute, Hungary (AEKD 02 A. Kotsarev Russian Research Centre "Kurchatov Institute", Institute of Nuclear Reactors, Russia (KI) 03 S. Nikonov Russian Research Centre "Kurchatov Instituteo', Institute of Nuclear Reactors, Russia (KI) 04 A. Pinegin Russian Research Centre "Kurchatov Institute", Institute of Nuclear Reactors, Russia (KI) 05 P. Gordienko Russian Research Centre "Kurchatov Institute", Institute of Nuclear Reactors, Russia (KI) 06 S. Kliem Helmholtz-TantrwDresden-Rossendorf, Germany (HZDR) 07 Y.Kozmenkov Helmholtz-ZentrumDresden-Rossendorf,Germany([ZDR) 08 E. Syrjalahti VTT Technical Research Centre of Finland (VTT) 09 G. Alechin FSU EDO Gidropress Podolsk, Russia (GP) l0 B. Kiss Budapest University of Technology and Economics, Institute of Nuclear Techniques, Hungary (BUTE) ll J. Hadek Nuclear Research Institute Rez, Czech Republic (NRD 12 I. Pasichnyk Gesellschaft fuer Anlagen und Reaktorsicherheit, Germany (GRS)

Guests 13 C. Parisi ENEA Roma,Italy 14 R. Gonzales University of Pisa, Department of Mechanical, Nuclear and Production Engineering, Italy 15 I. Gajev KTH Stockholm, Sweden 16 L. Sabotinov IRSN, France 17 E.Ivanov IRSN, France l8 S. Khodyachykh Siemens Erlangen, Germany LIST OF HAI{DOUTS

The following handouts of presentations were made available to the participants during the meeting, at least in electronic form. The corresponding PDF-files can be obtained from the chairman. tll C. Parisi, E. Negrenti, M. Sepielli: ENEA SIMING-LAB activities in the field of coupled codes simulations for the VVER technology I21 P. Gordienko, A. Kotsarev, M. Lizorkin: The new calculation of the 2"d Dynamic AER Benchmark * the ejection of a peripheral control rod in a WER-440 core t31 I.G. Petkevich, G.V. Alekhin, K.Yu.Kurakin: Calculation results of main steam line break with boundary conditions at the reactor using the coupled thermohydraulic computer code packages KORSAR/GP and TRAP-KS (in the framework of the international standard benchmark V1ffi0-CT2, exercise 2) t4l B. Kiss, A. Asz6di, I. Boros: Investigation of coolant mixing in VVER-440/213 RPV with SAS SST turbulent model t5l S. Nikonov, I. Pasichnyk, K. Velkov: Simulating UPTF C-3 experiment with advanced ATHLET t6l A. Pinegin, A. Kotsarev, M.Lizorkin: The statistical analysis of start-up experiment treatment on WER-1000 in hot zero power state I7l A. Keresztrlri, I. Panka: Coupling of codes for multiphysics analyses t8l M. Bendft, J. H6dek: The First Results of the 7s AER Dynamic Benchmark Problem. Solution with DYN3D and RELAP5-3D Codes t9l E. SyrjAhhti: Preliminary calculations of 7th AER Benchmark with APROS t10l S. Kliem, A. Grahn, B. Kiss, Y. Kozmenkov: Status of simulation of the 7ft AER Benchmark with DYN3D, ATHLET and ANSYS CFX AER Working Group E in 2011

Vladimfr Chrapdiak, VUJE, Inc., S lovakia, c hrapci ak@vuje. sk

The 16th meeting of the AER Working Group E "Physical Problems on Spent Fuel, Radwaste and Decommissioning of Nuclear Power Plants" was held in Modra - Harmonia, Slovakia, on 3-4May,20ll.

The meeting was focused on the following topics:

. storage and cask safety analyses r verification of nuclide composition calculation o PIE (Post Inadiations Experiments)

Total number of participants was l1 from 5 countries.

Mr. Becker informed about cask load planning in Germany. Mr. GerZa presented very good experience with dry storage at NPP Dukovany. Mr. Mikol65 informed about sensitivity analysis performed with the WIMS9 and WIMSlO code with 3 libraries based on JEFF 2.2 and JEFF 3.1. Mr. Svarnf published the uncertainties of MCNP and WIMS codes by methodology USLSTATS. Mr. Chrapdiak presented gammaspectrometric measurement of burnup in the inspection stand in Jaslovske Bohunice, results of CB6 Benchmark (defined by L. Markov6). MR. Zajac and Mr. Havluj presented 2 papers about preliminary evaluation of nuclide composition measurement ISTC 3958.

After block of presentations was general discussion with main topic: - a new measurement in RIAR. We have discussed about uncertainties of measurement. - IAEA meeting about Dual Cask Safety analysis.

On Wednesday we have planned the technical tour to Interim Spent Fuel Storage Facility in Jaslovske Bohunice. Unfortunately from security reason the technical tour was cancelled.

The next meeting will be held in Czech republic in April - May 2012. Agenda to 16th Meeting of AER Working Group E

Monday, May 2: arrival, dinner 18:30

Tuesday, May 3: meeting 10:00 opening l0:15 - 12:15 morning session, chairman: G. Hord6sy 10:15 - l0:45 A. Becker: Cask load planning and fuel pool management 10:45 - l1:15 J. Gerta: Experience with dry ISFSF in Dukovany l1:15- ll:45P.Mikola5:InfluencesofdifferentWIMSl0librariesonisotopeconcentrationsat bumup 1 l:45 - 12:15 J. Svamf: Remarks to uncertainties and Gd credit methodology

12:15 - 14:00 lunch 14:00 - 17:30 afternoon session, chairman: P. Mikol65 14:00 - 14:30 V. Chrapiiak at all.: Gamma spectrometric measurement ofbumup 14:30 - 15:00 V. Chrapdiak: Preliminary evaluation of BUC for WER-440 fuel (4.57%) l5:00 - 15:30 R. Z,ajac, V. Chrapdiak: Preliminary evaluation of ISTC 3958 measurement

15:30 - l6:00 coffee break 16:00' 16:30 F. Havluj: Preliminary evaluation of benchmark ISTC 3958 16:30 - 16:45 V. Chrapdiak: CB6 benchmark on VVER-440 final disposal 16:45 - l7:30 discussion (new IAEA activity - Dual cask) 18:00 dinner sponsored by WJE

Wednesday, May 4: visit of Interim storage in Jaslovsk6 Bohunice with inspection stend cancelled List of participants l) Jiif GerZa, NPP Dukovany,Czech Teyfa/e-m ail: +420561 103 I 43 / +420561 1049 59 / iiri. sena@,cez.cz

2) Franti5ek Havluj, UJV ReZ, Czech TeVfax/e-mall: +420 2 6617 3289/ +420 2 6617 23901 haf@,uiv.cz

3) Pavel Mikol65, SKODA JS, Plzen, Czech Tel/faxle-mail: +420 378042828/+420378042407/ pavel.mikolas(@skoda-js.cz

4) Jiif Svarnf, SKODA JS, Plzen, Czech

Tel/fax/e-m ail : + 42037 8M2828 / + 42037 8042407 / j i r i. sv arny@ sko da-j s. cz s) Sandor Patai-Szabo, Ts Enercon Ltd., Hungary

Tel/fax/e-m ail: 30 I 937 2628 / patai szabo. san dor@upcmai l. hu

6) Gribor Hord6sy, KFKI, Hungary Tel/fax/e-mail: +36 13922222 ext.3442/ + 36 I 3959293lhordos),@aeki.kfki.hu

7) Axel Becker, Studsvik Sandpower Tel/fax/e-mail: +49 40 3098 088 10/+49 40 3098088 88/ [email protected]

8) Silja Hiikkinen, VTT, Finland Tel/faxle-mail: +358 40 023 6898/+358 20 722 5000/ [email protected] e) Vladimfr Chrapdiak, VUJE a.s., Slovakia

T el/ fax/ e-mail: + 421 -33 - 599 I 3 12 I + 42 l -33 -599 I I 9 | / ch rapci ak@vuj e. s k l0) Radoslav Zajac, VUJE a.s, Slovakia TeVfax/e-mall: +421-33-5991316/ +421-33-5991191/ [email protected] l1) Juraj V6clav, UJD, Slovakia Tel/faxle-mail : Juraj.vaclav@uid. gov. sk 21th SYMPOSIUM of AER on WER Reactor Physics and Reactor Safety September 19 - 23, 2011, Dresden, Germany

Summary of

13th session of the AER Working Group F - "Spent Fuel Transmutations" and 4th meeting of INPRO Project RMI - "Meeting energy needs in the period of raw materials insufficiency during the 21st century"

held in Liblice, Czech Republic, April 26-29,2011.

Abstract.

lnformation about the progress in the problems spent fuel transmutation and future nuclear reactors development during the last years 2010 - 2011. lt is step by step clear that energy must be studied all together and the fact that the nuclear fission energy has the greatest resource basis must be used to keep long time stabilfty of supply and sustainability of development and this should be also decisive for the decisions what to do and what will be needed by society to keep contemporary high standard of living.

wgf11fi5.doc 6.9.2011 14:47 Page 1 of 7 All presentations are on the internet reachable through the command

ftp.//liblice2O 1 1 . uiv.cz together with some local information. Summarizing the meeting we would like to mention several basic presentations - overview of works concerning nuclear fuel resources, which are stating that there is enough resources (without uranium in oceans) to live with nuclear energy, supporting world mankind for more than thousand years. Such estimations are naturally out of any realistic social models - let us imagine, what was before thousand years. Using such ideas we silently suppose that this particular type of society, which is now in different degrees of development existing in our world will be during such long period and that the development will be without any hard breaks. - there was a model presentation of including nuclear energy into economy of developing countries and together with former presentations show differences with developed countries. lt is surely very good work, but facing contemporary European tendencies (especially in Germany) to refuse any nuclear power-stations, we see that even our society in Europe is more likely unpredictable and not using technological findings. - contradictions of EU supported program of fast helium cooled reactor with US supported program of high temperature reactor cooled by molten salts were described - it should be further clarified if the helium cooled reactor could support nonelectric applications and how construct heat exchanger to bring heating into the chemicalworkshops to produce for example hydrogen from water. - as a completely new element, two Japanese specialists visited Liblice session and presented their view on future development of traditional Oak-ridge molten salt reactor, having vision that it will at the end lead to use of thorium and to the safer fuel cycle. lt should be mentioned that their vision is now based on concrete idea how prepare in the first stage repetition of Oak-ridge reactor with contemporary materials and with contemporary safety and proliferation resistance demands in preparation. - current status of R&D on molten salt reactor and pyrochemical technology was thoroughly described and role of EC-EURATOM and GIF in coordination of it was analyzed. There still further improvement of fluoride volatility separation method, which is supposed to use for the first separation step of uranium from the solid spent nuclear fuel. - there was an attempt to start dialog with public and an hour discussion with the representative of ministry of environment of the Czech Republic were prepared. But at the last time l. Hlavac has apologize due to illness. Nevertheless discussion lead without concrete aim was misleading. lt is probably necessary to prepare such activity more thoroughly. Czech representatives of RAWRA (Radioactive Waste Repository Authority) presented their view on final disposal repository and there was no answer from them that natural process will lead to the reprocessing of spent fuel

wgfi lfs.doc 6.9.2011 14:47 Page 2 of 7 and that this completely changes their plans and constructions. lt time not to present people spent fuel as some wastes bus as new valuable raw material. It should be noted that meeting was before the Germany conclusion step by step close nuclear energy in their country. ln the last time there was also referendum in ltaly with the result not to continue with activities to use nuclear energy. On the other hand Poland parliament uniquely supported construction of the first nuclear power-station. We are speaking about it to support common activities with the aim to understand sources of people's fear from nuclear energy and evaluate how to overcome it. Conclusion of meeting was approved based on IAEA suggestion and is in the Enclosure 2. Enclosure 1 is complete program. To simplifiT orientation in the problems of RMI group it was prepared CD (1.6M8) collecting works of different authors with relation to the work RMI group. We would like to underline set of US works, concerning hydrogen production and necessity to clarify how to connect it nuclear source of heating to drive the process. This will be also answer on IAEA INPRo initiative speaking about nonelectric application of nuclear energy.

It was recommended during the meeting to intensify works and prepare something like "call for cooperation" to be able to understand problems of future demands and challenges in the connection with the coming raw materials deficiencies and growing of population and energy needs. Potential lefter shoutd be distributed via official channels of IAEA and lNPRo. see Enclosure 2.

Next meeting is supposed in the term April 10-l 3, 2012 at the same place in the Liblice chateau.

wgt11fis.doc 6.9.2011 14:47 Page 3 of 7 Enclosure 1.

Technical Program 73th ression of the AER Working Gtoup F - 'Spent Fuel Transmubffions' and 4st meeting of IIIPRO Prcject RMI - "Meeting enetgy needs in the prid of raw materials insuffici,enq during the 2 tst cenfrny " Liblice, Czech Republic, April26 - 3q 2UL

Tuesday, April 26 12.30 - 13.30 L u n c h (if ordered). - 14.00 Travel to Liblice, accommodation. 14.00 - 18.00 lndividual discussions about energy and evaluation of internet materia /s an d possibility to predict development. Evaluation of US works. published on www.doe.oov. Do we have really insufficiencies in raw materials and how to evaluate them? lnformation about the future fasks within INPRO in the international cooperation. National energy strategies and its technical support, how prepare data to calculate prediction. 18.30 - 20.00 Dinner. 20.00 - 21.30 Personal discusslons.

Wednesday, April2T 07.30 - 08.4s Breakfast. 09.00 - 09.05 Formal introductory information. 09.05 - 09.30 Lelek: Development during the last year. About the differences in US and EU programs. (Published results on hydrogen program and G4 cooperation are calling for sequenced evaluation by different specialists. Demonstration and possibility to copy literature.) 09.30 - 10.00 Hron : Conclusions of the SPHINX (SPent Hot fuel lncineration by Neutron fluX) proiect 10.00 - 10.30 Busurin: Revised draft term of reference INPRO Collaborative Project (IAEA-INPRO team) Meeting energy needs in the period of raw materials insufficiency durino the 21st centurv RMl. 10.30 - 1 1.00 Gaqarinskii: Fuel resources for nuclear energy. Current status and trends. 11.04 - 11.20 Coffee Break. 11.20 - 11.55 Tsibulskiy: Probable nuclear energy development scenarios for rapidly develooino economies. 11.55 - 12.10 Ail Energy problem from different points of vrbws - what could be consequences due to wrcng planning - possib/e consequences of low enemv supplv in some develooino counties. 12.30 - 13.50 Lunch. 14.00 - 14.45 Furukawa: Early Realization of Th Fuel-Cycle by FUJI Molten Salt Reactor Svstem

wgfl1fis.doc 6.9.2011 14:47 Page 4 of 7 14.45 - 15.'15 Uhlir (presented Current status of R&D on MSR and pyrochemicaltechnology by Hron): under coordination of EC-EURATOM, GIF and IAEA 15.15 - 15.35 Bican: Current Status in Experimental Molten Salt Loops. 15.35 - 15.55 Skarohlid: Current Progress in Fluoride Volatility Process in NRI Rez. 15.55 - 16.15 Straka, Electrochemical separation of selected fissile material and fission Szatm6ry: products representatives in molten fluoride melts. 16.15 - 16.40 Coffee Break. 16.40 - 16.55 Slov6k The Czech Geological Repository - impacts on the strategy of the fuel cvcle back end oublic views.

16.55 - 17.55 Hlavac lMtat could be economical and sociological consequences of the (apologized): aming raw mateials cisis. {dirccted discussion with sociologist Discussion with about interaction of technological and raw material situation with professional people as sociological grcups - developed and developing world; sociologist and different economical consequences based on the transport economist limitations; what are fhe reasons for the people to move.] 17.55 - 18.05 Lelek: Preliminary conclusions and ..

18.10 - 18.45 Personal discussion, Cultural program (under considerations)

18.45 - 20.00 Dinner 19.00 - 21.30 Evening program - discussion continuation.

Thurcday, April 28 07.00 - 07,30 Breakfast 09.00 - 09.15 Wagner: First experiments with big uranium set-up Kvinta irradiated by deuterons 09.15 - 09.30 Svoboda: Comparison of neutron production on lead-uranium set-up irradiated bv protons and deuterons with different energies 09.30 - 09.40 Suchop5r: Simulations of new uranium set-up l(/INTA used on Nuclotron accelerator at Dubna 09.40 - 09.50 Stef6nik: Determination of Neutron Room Background at the NPI Cyclotron u-120M 09.50 - 10.10 Daiilek, WER-440 with IMF - safety analysis of control rod ejection Strmenskf, Zalac, Majerdik 10.15 - 10.30 Zaiac, Daiilek Advanced Combined Fuel Assembly for WER-440 10.30 - 10.45 Daiilek ALLEGRO - status and ambitions Long time fuel cycle simulation. 10.45 - 1 1.05 Brolly:

11.05 - 11.10 Coff ee B e a k (also durinq the discussion). 11.10 - 12.00 all Discussion about future common cooperation about the technical and political development of energy problem and do we have any idea about expectations for due to future limitations in raw materials supply. Should we formulate wort

ls there a European energy policy - technical and political - what should be exoeded in future? 12.00 - 12.30 all or chosen Reserve for specific presentation or discussion.

wgfllfiS.doc 6.9.2011 14:47 Page 5 of 7 oroup 12.00 - 12.30 M e m o ra n d u m f o rm u latio n. 12.30 - 13.30 Lunch. 14.00 - 19.00 Excudon through the castle and park,lecturc about history, contemporary use of the locality.

http : //www. z a m e k-l i b I ice. cz all Personal discussions during excursions. 20.00 - Closing internal discussion

Friday, April 29 07.00 - 07.30 Breakfast. 08.00 - 08.30 Closing discussi on. 09.00 - 10.30 Dataforforecast p r e p a r a t io n ( t e s t c a lc u la t io n s ) 10.30 - 10.50 Coff ee Break up to lndividual negotiation in NRI Rez and Saturday, Aprill0 in Technical University, Prague.

List of participants 2011.

number of country institution names oarticinants

4 Czech Reo. UJF AV CR Rez Wagner, Svoboda, Suchopir, Stetinit 4 Czech Rep. UJV Rez Lelek, Hron, Mikisek, Juricek 3 Gzech Rep. UJV Rez Straka, Skarohlid, Bican 1 Czech Rep. UJV Rez Hrehor 1 Gzech Rep. FJFI CVUT Kobylka I Czech Reo. Eneroowzkum Matal 3 Gzech Reo. SURAO Kaplin, Slovik, SumberovS 1 Czech Reo. JE Dukovanv Bajgl 1 Czech Reo. Ministrv of the Environment Hlavid (appologized at the last moment) 2 Slovakia VUJE Trnava Dafflek, Zaiac 2 Slovakia Ustavu anorganickej chemie Korenko, Simko SAV Bratislava 2 Austria IAEA, INPRO team Busurin, Belova 2 Hunqarv KFKI. Technical Universitv A. Brolly, Magdolna MISKOLCZY

2 Russia RRC "Kurchatov institute' A. Gagarinskiy, V. Tsibulskiy 2 Japan Thorium Tech Solution K. Furukawa, M. Furukawa

wail11f5.doc 6.9.2011 14:47 Page 6 of 7 Enclosure 2.

Proposal of the final Minutes of RMI meeting (25-29 of April, 201,1, Liblice, Czech Republic)

New Terms of Reference of INPRO Collaborative Project 66RMf" (Meeting energy needs in the period of raw materials insufficiency during the 21st century) was presented by Y.Busurin, IAEA INPRO group member and was discussed by meeting participants from the Czech Republic, Russian Federation, Slovakia and Hungary. It was agreed that RMI may significantly contribute to the INPRO "White paper" elaboration of Global vision of sustainable Nuclear Energy development in the 21-st Century in a new context of New ToR of RMI and in particular: - Nuclear Energy contribution to the non-electrical applications with a special emphasis on motor fuel production, hydrogen production and water desalination; - Availability of resources for Nuclear Energy development in the 21-st Century (uranium, thorium, construction and other raw materials and human resources);

- Nuclear materials flows scenarios in different regions and multinational Nuclear fuel cycle centres considerations for newcomers (institutional, economics, environment and security issues); - NPP accident consequences mitigation issues for energy supply to the countries' economics. - Nuclear Enerry contribution to the economics of fast developing countries and regions;

It was recommended that IAEA INPRO group will develop a Draft content of "White paper" with identification of additional areas where RMI contributions might be welcomed. It was recommended to inform current Steering committee about the meeting and its conclusions. It was agreed to conduct Consultancy meeting in Vienna at the first week of October to finalize RMI ToR with identification of all participants' contributions and time schedule for Project implementation. RMI group is analyzing period of energy needs and possible nuclear energy application in the nearest time and that is why it is of primary interest for the planning of sustainable future.

Liblice, April 29,2011

wgfl 1tis.doc 6.9.201 1 14:47 PageT oI7 SUMMARY ON THE ACTIVITY OF WORKING GROUP G IN 2OI1

Dr. Attila Asz6di, Coordinator of Working Group G Budapest University of Technology and Economics Institute of Nuclear Techniques I I 1l Budapest, Miiegyetem rkp. 9., Hungary aszodi@reak bme.hu

The AER Working G meeting was held on 2-3 June 2011 in Balatonftired (Hungary) together with Working Group C. 28 participants from 6 countries and from 11 organizations were participating on the meeting. All together 18 presentations were shown, l0 presentations in the session of WG C, and 8 presentations in the frame of the working group G.

The program of the Working Group G meeting was the following.

1. Toppila, Timo (FORTUM) CFD Simulations Supporting Transition to 4-batch Loading in Loviisa 2. Szij6rt6, Rita (BME) Study of Natural Convection Around a Vertical Heated Rod 3. T6th, Sdndor (BME) Modelling of Thermal Mixing in a T-junction 4. Hohne, Tomas (HZDR) OECD/NEA & Vattenfall T-Junction Benchmark Exercise 5. Kubadka, J6n (VUJE) CFD Model of Passive Autocatalytic Hydrogen Recombinator 6. Kudryavtsev, O. (GIDROPRESS) Analysis of Steam Flow Rate Distribution Nonuninformity Through Distribution Perforated Sheet Holes in Steam Generator

7. Posysaev, Dmitry (GIDROPRESS) Results of WER-440 Fuel Assembly Head Benchmark

8. Kiss, B6la (BME) lnvestigation of Coolant Mixing in WER-44O12L3 RPV with SAS SST Turbulence Model

The presentations are available by the working group coordinator. More detailed summary is given in the presentation No. 6.1. on the Symposium 2011. AER WORKING GROT]P H ACTIVITY IN 2011

L.Shishkov. National Research Centre

ABSTRACT

The paper reports on Moscow meeting of working group H, which was held in "Kurchatov Institute", June 28-29. The meeting was attended by 26 physicists from 13 institutions of Hungary, Czechia, Slovakia, Germany, Russia and the Ukraine. Name-list of the participants and titles of their presentations may be found in the paper, which also gives brief information on the designation of presentations.

The meeting of working group H "Elaboration of methodology of calculating the core design engineering factors" was held in Moscow at NRC "Kurchatov Institute" on June 28- 29,2011. The meeting was attendedby 26 experts from 13 institutions of 6 member-states. Please, find below the name list of the participants and the titles of presentations.

Name List Name, Surname Organization Country l. Josef Bajgl (EZ, a.s.,Dukovany NPP Czech Republic 2. Jiri Svarny SKODA JS a.s. Czech Republic 3. Martin Sa5ek Srone JS, a.s. Czech Republic 4. Vladislav Stary Nuclear Research Institut, REZ, plc Czech republic 5. Michal Sedansky VUJE,Inc. Slovak republic 6. Mikuki5 Bendat Urad Jadroveho Dozoru of Slovak Slovak republic Republic 7. Juraj Simko Slovenskd Elektr6rne - ENEL Slovak republic 8. Rohde Ulrich HZDR Institute of Safety Research Germany 9. Keresztriri Andrrls MTA KFKI Atomic Energy Research Hungary Institute 10. Ievgen Bilodid SSTC N&RS Ukraine I l. Shishkov Lev NRC Russia 12. Lizorkin Mikhail NRC Russia 13. Saprykin Vasiliy NRC Russia 14. Tsyganov Sergey NRC Russia 15. Dementiev Viatcheslav NRC Russia 16. Kalinushkin Andrey NRC Russia 17. Sidorenko Vladimir NRC Russia 18. OleksyukDmitry NRC Russia 19. Oleinik Dmitry NRC Russia 20. Kurakin Konstantin OKB ''GIDROPRESS'' Russia 21. Ustinov A. OKB'GIDROPRESS" Russia 22. Podshybyakin A. OKB'GIDROPRESS' Russia 23. Petkevich I. OKB ''GIDROPRESS'' Russia 24. Bykov Alexandr JSC "IF SNIIP Atom" Russia 25. Kryanev A. MIPI Russia 26. Romodanov V. MIPI Russia

List of presentations I L. Shishkov, V. Dementyev, D. Oleksuk . 2 J. Svarn;f, M. Sa5ek < A. Keresztriri . L. Shishkov, V. Dementyev <. K. Kurakin < M. Sa5ek, J. Svarnj M. Sedansky A. Kuzhil, E. Kulik, T.Makarova, S. Padun A. Bykov, V. Osipov t0 J. Bajgl ll J. Bajgl 12 V. Milto, A. Kalinushkin l3 M. Bencat > t4 D. Oleinik l5 A. Kryanev, G. Lukin, D. Udumyan l6 V. Romodanov The presentations from the list can be split into four topical groups: I Group I includes presentation l, which makes an effort to elaborate the draft final document, which will conclude the meeting of Group H in case of optimal scenario. The document is compiled in the form of presenting the main provisions of evaluation method of margin factors. Some of the provisions are presented on the form of several alternative solutions. The discussion of the presentation did not bring the participants to a common opinion, and it was decided to continue the discussion. U Group 2 includes presentations 2,3, 4, 5 and 12. The presentations discuss certain points in the methods of evaluating engineering margin factors. For example, assurance of required probability of meeting design limitations, and assuring the confidence in providing such probability; requirements for the characteristics of retrievals of comparisons between calculated and measured data, etc., the presentations discuss the list of parameters to be limited by design, and the of conditions under which the limitations shall be satisfied.

III Group 3 includes presentations 2,6,7,8,9, 10, 11 and 13, which discuss the application effects of margin factors and the prospects of applying the margin factors for fuel cycles of active power plants. The presentations compare the methodological uncertainties of various computer codes, assess the ICIS uncertainties, discuss and evaluate the systematic uncertainties typical for certain reactor cores. ry Group 4 includes presentations 14, 15 and 16 that are not associated directly with engineering margin factors, however, they can be useful in determination or checking the margin factors. For example, presentation 14 discusses the improvement of code MCU to be applied in calculation of sensitivity factors of power density parameters to deviations in technology factors; presentation 15 discusses a mathematical method that allows realization of effective methods of multiple valued functions for interpolation and extrapolation. Finally, presentation 16 suggests a simple and cheap method of on-site experimental non-destructive control of fuel content and enrichment.

Each presentation is available on computer as PPT file.