Working Party on Nuclear Criticality Safety

Transcription

1 For Official Use NEA/NSC/WPNCS/DOC(2008)8 NEA/NSC/WPNCS/DOC(2008)8 For Official Use Organisation de Coopération et de Développement Économiques Organisation for Economic Co-operation and Development 18-Dec-2008 English - Or. English NUCLEAR ENERGY AGENCY NUCLEAR SCIENCE COMMITTEE Working Party on Nuclear Criticality Safety Seventeenth Meeting of the Expert Group on Burn-up Credit Criticality SUMMARY RECORD 22 September 2008 NEA Offices Issy-les-Moulineaux, France English - Or. English Yolanda RUGAMA Yolanda.RUGAMA@oecd.org +33 (0) JT Document complet disponible sur OLIS dans son format d'origine Complete document available on OLIS in its original format

2 Working Party on Nuclear Criticality Safety Seventeenth Meeting of the Expert Group on Burn-up Credit Criticality 22 September 2008 Issy-les-Moulineaux, France SUMMARY RECORD 1. Welcome to the BUC EG meeting and practical information (M. Brady Raap) Twenty-nine participants (see the list of participants provided in Annex 1) attended the meeting. The chair, Brady Raap, opened the meeting and welcomed the participants. The Phase II-C report was distributed to the participants. 2. Approval of the agenda and review of actions from the previous meeting (Y. Rugama) The proposed agenda (see Annex 3) was adopted after the inclusion of a sub-item, Criticality analysis for the new type of Gd fuel at the Paks NPP, under item 7. Rugama reviewed the progress of actions taken since the previous meetings. The first two actions will be discussed under agenda item 5. In addition, participants to the Phase II-E report were asked to send their contributions to Neuber before the end of November Neuber acknowledged the contributions and will discuss the preliminary analyses of the Phase II-E report under agenda item 6. Finally, an action involving Brady Raap and Wagner concerned the distribution of specifications for the Phase VII report. Parks will provide a preliminary version (Annex 4) for discussion under agenda item Approval of the summary record The summary record of the previous meeting was approved without modification. 4. BUC national programs Riffard reported on the progress of the collaboration between industry (EDF, AREVA), and state-run organizations (CEA,IRSN) to produce a validated burn-up credit methodology for France. The French burn-up credit profile analysis from experimental data has reached the final stage. The CRISTAL package has been updated for MOX applications. The European library JEFF3.1 (its corrections are available at the NEA website) is being used for burn-up credit applications and is based on the results obtained. Improvements have already been proposed for future versions of the library. The results of the sensitivity analyses of the Takahama-3 data available in SFCOMPO were briefly discussed. De Hemptinne discussed the Belgian activities related to burn-up credit. Tractebel has allocated funding for their participation in the Phase II-E benchmark exercise and they expect to be able to participate in expert group activities related to geological disposal applications. Hordosy reported on developments in burn-up credit studies in Hungary. The criticality safety analysis for the new type of Gd fuel for the Paks Nuclear Power Plant was discussed extensively under agenda item 7. 2

3 Chrapciak presented the status of the validation of criticality and inventory calculations using SCALE5.1 in the Slovak Republic. Transport casks using BUC are in the process of being licensed. Because of the lack of isotopic composition data for VVER reactors, only actinides have been used in the validation. Markova informed the participants of advances made by burn-up credit studies in the Czech Republic. She presented the program launched to increase NPPs efficiency. The NRI analysts corrected errors in the recent #2670 ISTC VVER PIE report and made significant improvements to the benchmark specification based on the ISTC #2670 PIE project (Phase VI). New calculations have been performed to implement the corrections; the results will be discussed under agenda item 7. Gulliford briefly discussed the already existing activities on BUC, which are ongoing but stated that no new activity associated with BUC has been launched in the UK. He highlighted the UK s interest in participating in the group activities related to legacy waste. Neuber and Gmal informed the participants about improvements to inventory calculation methodology using the data available in Germany. The application of burn-up credit to dry storage and transport is under study. The evaluation of PIE data from various sources is being used for the validation of the following codes: CASMO-IV and TRITON (SCALE 5.1) and MCNP. Neuber has also applied the missing-data evaluation method to isotopic correction factors in criticality safety analyses for different BWR designs and pools. Parks discussed the status of burn-up credit implementation in ORNL in the USA. ORNL continues to work on a technical basis justification for implementation of burn-up credit. In particular, ORNL is working to obtain and evaluate additional assay and critical experiment data from domestic and international sources in order to ascertain their usefulness in supporting validation of burn-up credit analyses. An evaluation of the HTC experiments performed by IRSN was completed and a report on the assessment will soon be published by the Nuclear Regulatory Commission. Through a contractual agreement with AREVA and IRSN, these proprietary HTC experiments are now available in the US for domestic licensing for storage, transport, and disposal. Domestic efforts to obtain and perform assay data for PWR spent fuel are being planned and work to obtain the samples has been initiated. Updated analyses with the latest version of SCALE are being carried out with all the publicly available assay data. Suyama reported on the implementation of burn-up credit in Japan. He presented the SWAT3.1 system (an integrated burn-up calculation code system), the methodology used and its nuclear data libraries. Suyama also gave an update on the progress of studies organised in Japan on the end effect. Kolbe presented the BUC applications performed in Switzerland. BUC has been limited to PWR wet storage pools and based on the BOXER 2-D transport code. The short-term goal is to upgrade the analysis methodology to CASMO-4 for depletion calculations and MCNPX for criticality analyses. The validation of the former continues in parallel, and is mainly based on Swiss spent fuel data. Conde informed the participants that the Spanish regulatory body (CSN) has licensed an Interim Spent Fuel Storage Installation at the José Cabrera NPP site (160 Mw, ), where spent fuel will be loaded in dual purpose casks (storage HI-STORM 100 / transportation HI-STAR 100) with implementation of burn-up credit for the transportation case only. The spent fuel of the plant, permanently shut down in April 2006, will be transferred from the original spent fuel pool to an open-air dedicated area located inside the plant site for the storage of the casks. The first cask will be loaded before the end of

4 Mennerdahl observed that there has not been any significant change in the nuclear energy situation in Sweden during the past year. A new licensing authority, SSM (Strålskyddsmyndigheten or the Swedish Radiation Safety Authority), was created on 1 July 2008 from a merger of the old authorities for nuclear safety SKI (Statens Kärnkraftinspektion or the Swedish Nuclear Power Inspectorate) and for radiation protection SSI (Statens Strålskyddsinstitut or the Swedish Radiation Protection Institute). 5. Status of Report summarising the results of the Expert group (M. Brady Raap) Since the last meeting not much progress has been made on this summary report: preliminary draft versions of the PWR and BWR chapters have not been made available yet. The status of each group s paper is as follows: Application to PWR VVER-design fuel: Markova and Hordosy A draft version will be sent to the NEA secretariat for posting on the EG website. Application to PWR-UOX: DeHart and Neuber Neuber reported that not much progress has been achieved since the last year. He agreed to lead the PWR-UOX chapter and will provide a draft version before June Application to BWR: Conde and Okuno Conde agreed to prepare the draft version of the BWR contribution for December Validating burn-up credit criticality: Gulliford and Santamarina Gulliford is expecting new data from the CEA to complete the chapter. The participants agreed to have the contributions completed for review by the June The final report after internal BUC review should be available by the end of October The draft versions will be uploaded on the EG protected website area for the reviewing process. Brady Raap added that the report should be completed before the next meeting, otherwise the activity would be abandoned. 6. Status of Phase II-E benchmark (Neuber) Neuber presented the preliminary analyses of the Phase II-E report. He reminded the participants that the Phase II-E benchmark exercise consists of the following sets of calculations: Determination of the neutron multiplication factor K eff of the cask configuration for the two axial burn-up profiles and at least nine different Control Rod (CR) insertion depths (cf. section 2.2.1). Determination of the neutron multiplication factor K eff of the cask configuration for the uniform distributions of the average burn-up values of 30 MWd/kg U and 50 MWd/kg U and for the nine CR insertion depths. 4

5 Neuber commented that since the last meeting he has received new contributions from Belgium, France (IRSN/AREVA) and the UK. He acknowledged the participants for their inputs. Neuber concluded that the asymmetry of axial burn-up profiles decreases with increasing average burn-up, and that the axial end effect at any given asymmetry increases with increasing average burn-up. He added that this is true for any given control rod insertion depth during irradiation of the fuel, but that the amount of the end effect varies significantly with the control rod insertion depth. This variation is correlated to the variation of the axial fission density distribution which is impacted by the asymmetry of the axial burn-up profile and depends on the average burn-up of the profile. Neuber proposed to investigate the possibility of performing an analysis than of Phase II-C, but with MOX fuel instead of UOX fuel. With regard to the question about the need to have detailed axial burn-up profiles, he replied that he could provide the profiles.the participants agreed to discuss the proposal at the next meeting. 7. Status of Phase VI benchmark, project ISTC 2670T (Chrapciak, Havluj, Hordosy) The analyses of data from the ISTC project N o 2670 funded by the U.S. DOE (managed by the Lawrence Livermore National Laboratory (LLNL) and the Research Institute of Atomic Reactors (RIAR)) is ongoing. The first results were presented by Chrapciak and Havluj. Markova and Havluj will lead the activity and will compile the results and perform the analyses. A benchmark has been prepared to simulate the operational history of the FA design examined in the VVER #2670 ISTC project. The benchmark specification is based on the #2670 ISTC VVER-440 PIE data which were reasonably simplified. The core pre-history/position characteristics were provided by the Russian experimentalist and corrections to the original reports were included. Two different calculation roots were used (based on the irradiation history/measured burn-up) and the results showed a slight difference. From the analyses of the results presented, it can be concluded that data from reactor operations are missing and that both experimental data and the evaluation of the project showed inaccuracies. Chrapiack presented the real burn-up radial profile in the VVER-440 assembly (3.6% enrichment). The codes used were SCALE 5.1 and PERMAK-3D. Comparisons of diagonal relative burn-up distributions for the ideal case (SCALE 5.1) and the real case at various locations (11, 19, 21 and 41) were discussed. Havluj discussed the results and evaluation of the benchmark exercise. He highlighted difficulties various participants encountered in estimating the burn-up through the burn-up indicator provided in the original reports ( 148 Nd). The isotopic inventory calculated by the participants was compared. He concluded that the participants agreed with the experimental values for most of the isotopes. However, correction factors were needed for other isotopes (e.g. 237 Np, 151 Eu). With regard to Meraj s question about the discrepancy of results for the Eu and Tc isotopes, Havluj replied that the difference could be explained by the quality of the experimental values and the uncertainty of the nuclear data. Suyama and Riffard suggested using another burn-up indicator, for instance 145 Nd or 148 Nd. Markova agreed to complete the analyses of the current data before proposing any dates for the final publication. 5

6 Hordosy presented the criticality safety analysis for the new type of Gd fuel that will be uploaded at the Paks NPP. He described the methodology developed for the criticality safety assessment in detail. The main distinctive figure of this fuel is the use of three gadolinium fuel pins instead of the six fuel pins commonly used. Because of the limited amount of experimental data available for this type of fuel, uncertainty analyses were carried out. The cross-section error and covariance matrices used at the transport calculations (MCNP) were extracted from the COVARX library (SCALE 5.1). The Gd crosssection was modified conservatively and isotopic correction factors were used. Hordosy concluded that the new fuel should meet the sub-criticality conditions required. 8. Future Activities 9. Adjourn Status of the ORNL proposal for a benchmark on BUC waste disposal related issues (C. Parks) Parks presented the ORNL proposal for a benchmark exercise on BUC waste disposal issues (Annex 4). The proposal was extensively discussed and it was agreed that some small changes be made to the first exercise. The exercise will be limited to PWR fuel in the cask model already used in past BUC benchmark exercises. The participants agreed to limit the calculations to 28 steps (105 years) and one initial enrichment (wt% 235 U)/burn-up (GWd/MTU) combination, 4.5/50. The list of isotopes on the proposal will be reviewed by the participants, who have planned to send their comments before the end of October Parks/Wagner agreed to collect the comments from the participants and send a new version of the proposal by November The list of participants/organisations is as follows: CEA, BFS,IRSN,Mennerdahl (2 libraries: ENDF, JEFF ), JNES, GRS, PSI, AEKI, TVO, NRI,Areva-France, Areva- Germany, Gulliford (2 codes: FISPIN, WIMS), CSN, NRC, ORNL. Benchmark exercise similar to Phase II-C but for MOX fuel (J-C. Neuber) Neuber suggested discussing the possibility of organising a similar benchmark exercise as the one organised for Phase II-C (end effect + CR insertation), but with MOX fuel. He will provide the axial profile as well as the specifications. The proposal has been very well-received and several participants showed an interest in contributing to this new activity. However, it was agreed that Neuber complete Phase II- E before sending the specifications for this new exercise. Neuber agreed to present the specifications for this new exercise at the next meeting. Brady-Raap proposed to hold the next meeting in conjunction with the IAEA Technical Meeting on BUC that will be held in Spain. The participants agreed to meet the Monday before the IAEA TM in Spain. Rugama will provide the delegates with the final location and dates. 6

7 ANNEX 1 Country and Name Establishment BELGIUM Ms. Gwendoline DE HEMPTINNE TRACTEBEL gwendoline.dehemptinne@tractebel.com CZECH REPUBLIC Frantisek HAVLUJ PRAGUEREZ haf@nri.cz Dr. Ludmila MARKOVA PRAGUEREZ mar@nri.cz FINLAND Mr. Anssu RANTA-AHO TVO anssu.ranta-aho@tvo.fi FRANCE Mr; Stephane EVO IRSN stephane.evo@irsn.fr Ms. Ludyvine JUTIER IRSN ludyvine.jutier@irsn.fr Mr. Yi-Kang LEE SACL-SERMA yklee@cea.fr Mr. Vincent LEGER TNINT vincent.leger1@areva.com Ms. Cecile RIFFARD CADARACHE cecile.riffard@cea.fr Ms. Veronique ROUYER IRSN veronique.rouyer@irsn.fr GERMANY Dr. Robert KILGER GRS robert.kilger@grs.de Dr. Jens-Christian NEUBER AREVA-OFF jens-christian.neuber@areva.com Mr. Ingo REICHE BFSSALZ ireiche@bfs.de HUNGARY Dr. Gabor HORDOSY MAGYARATOM hordosy@aeki.kfki.hu JAPAN Mr. Kouji HIRAIWA TOSHIBA kouji.hiraiwa@toshiba.co.jp Mr Yoshinori MIYOSHI JAEATOK-SC miyoshi.yoshinori@jaea.go.jp Dr. Kenya SUYAMA MEXT suyama@mext.go.jp Dr. Toru YAMAMOTO JNES yamamoto-toru@jnes.go.jp SLOVAK REPUBLIC Dr. Vladimir CHRAPCIAK VUJETRNAVA chrapciak@vuje.sk SPAIN Ms. Consuelo ALEJANO MONGE CSN cam@csn.es Mr. Jose Manuel CONDE LOPEZ CSN jmcl@csn.es Dr Pedro ORTEGO SEAMADRID p.ortego@seaingenieria.es 7

8 Country and Name Establishment SWEDEN Mr. Dennis MENNERDAHL EMSYSTEMS SWITZERLAND Dr. Edwin E. KOLBE PSI UNITED KINGDOM Dr. Jim GULLIFORD NNL-HAR UNITED STATES OF AMERICA Dr. Michaele C. BRADY RAAP PNLWA Dr. Germina ILAS ORNL Dr. Richard MCKNIGHT ANL Dr. Cecil V. PARKS ORNL Meraj RAHIMI NRCROCK International Organisations Dr. Yolanda RUGAMA SAEZ NEADB 8

9 ANNEX 2 LIST OF ACTIONS FROM THE 17TH MEETING OF THE EXPERT GROUP ON BURN-UP CREDIT Action number Action holders Description of the action Due dates Have the draft version of the Buc Conde reviewing by December 2008 BWR chapter for the lessons December learned report completed for 2008 Have the draft version of the Buc Neuber PWR chapter for the lessons learned report completed for June 2009 reviewing by June 2009 Buc CEA(Santamarina) Send comments to Gulliford about the chapter Validating BUC criticality of the lessons learned report January 2009 Buc Buc All Parks, Wagner The contacts for the lessons learned papers are: Send comments and list of isotopes for the Phase VII benchmark exercise Include comments from participants to the specifications Phase VII L. Markova and G. Hordosy: Application to PWR VVER-design fuel G. O Connor and Hong: Application to PWR-MOX fuel J-C. Neuber: Application to PWR-UOX End October 2008 November 2008 J-M. Conde leading the report and H. Okuno for reviewing: Application to BWR J. Gulliford and A. Santamarina: Validating burn-up credit criticality 9

10 ANNEX 3 Nuclear Energy Agency Nuclear Science Committee Working Party on Nuclear Criticality Safety THE SEVENTEENTH MEETING OF THE EXPERT GROUP ON BURN-UP CREDIT CRITICALITY NEA Offices, Issy-les-Moulineaux, France 22 September :00 AM to 5:30 PM PROPOSED AGENDA 1. Welcome to the BUC EG meeting and practicalities M. Brady Raap 2. Approval of the Agenda and review of actions from previous Y. Rugama meeting 3. Approval of the summary records All 4. Status of BUC Activities in the Countries and International Organisations 5. Status of Report summarizing the results of the Expert group: PWR, BWR, MOX, VVER and Validation All Leaders for each reactor system (see table on page 2) 6. Status of Phase II-E benchmark J-C. Neuber 7. Status of Phase VI benchmark, project ISTC 2670T Real Burnup profile in the VVER-440 assembly (3.6% enrichment) VVER PIE database for EG ADSNF and ISTC #2670 benchmark evaluation 8. Future Activities Status of the ORNL proposal for a benchmark on BUC waste disposal related issues 9. Adjourn V. Chrapciak F. Havluj ORNL 10

11 ANNEX 4 Proposed Specification for Phase VII Benchmark (Submitted to the NEA Expert Group on Burnup Credit for their consideration) September 22, 2008 UO 2 Fuel: Study of spent fuel compositions for long-term disposal John C. Wagner (ORNL, USA) Georgeta Radulescu (ORNL, USA) Mikey Brady-Raap (PNNL, USA) 1. INTRODUCTION The concept of taking credit for the reduction in reactivity due to fuel burnup is commonly referred to as burnup credit. The reduction in reactivity that occurs with fuel burnup is due to the change in concentration (net reduction) of fissile nuclides and the production of actinide and fission-product neutron absorbers. After spent nuclear fuel (SNF) is discharged from a reactor, the reactivity continues to vary as a function of time due to the decay of unstable isotopes. Burnup credit analysis for storage and transport considers timeframes that are extremely short (typically less than 100 years), as compared with the timeframe of interest to long-term disposal (e.g., 10,000 years after closure in the U.S.). This benchmark proposes to study the ability of relevant computer codes and associated nuclear data to predict spent fuel isotopic compositions and corresponding k eff values in a cask configuration over the time duration relevant to SNF disposal. It is recognized that the benefits of this proposed benchmark are largely confined to revealing differences in nuclear data for decay constants (half-lives, branching fractions), which are widely considered to be well known. However, the results of this exercise may serve to reveal differences in international nuclear data sets and/or improved understanding and confidence in our ability to predict k eff for timeframes relevant to long-term disposal of SNF. The benchmark is divided into two sets of calculations: Decay calculations for provided PWR and BWR UO 2 discharged fuel compositions Criticality (k eff ) calculations for a representative cask model. Participants are requested to perform decay calculations for PWR and BWR fuel compositions and criticality calculations for PWR fuel in the OECD cask model for a fairly large number of post-irradiation time steps (~30), out to 100,000 years. Although it is acknowledged that the physical condition of the 11

12 fuel will change over such a long time period, there is interest in the change in isotopic compositions over this duration, as well as interest in the relative behavior of k eff over this duration. Analysis of the results will involve comparison of the participant s isotopic compositions and k eff values as a function of time. 2. DECAY CALCULATIONS Utilizing the discharge compositions of PWR and BWR assemblies for the actinide and fission product burnup credit nuclides provided in Table 1 (Set 2), participants are requested to perform decay calculations and report isotopic compositions for the times listed in Table 2. The discharge isotopic compositions of fuel, in units of atom/barn cm, will be provided to the participants as text files attached to the benchmark specifications. The isotopic composition values will be provided for a representative PWR and a representative BWR assembly and two relevant initial enrichment (wt% 235 U)/burnup (GWd/MTU) combinations, 3.0/30 and 4.5/50. Table 1: Nuclides for which compositions are to be calculated and provided Set 1: Actinide-only burnup credit nuclides (10 total) 233 U, 234 U, 235 U, 238 U, 238 Pu, 239 Pu, 240 Pu, 241 Pu, 242 Pu, and 241 Am Set 2: Actinide + fission product burnup credit nuclides (29 total) 233 U, 234 U, 235 U, 236 U, 238 U, 237 Np, 238 Pu, 239 Pu, 240 Pu, 241 Pu, 242 Pu, 241 Am, 242m Am a, 243 Am, 95 Mo, 99 Tc, 101 Ru, 103 Rh, 109 Ag, 143 Nd, 145 Nd, 147 Sm, 149 Sm, 150 Sm, 151 Sm, 152 Sm, 151 Eu, 153 Eu, and 155 Gd Table 2: Times for calculating and reporting isotopic compositions Time case number Time (y) Time case number Time (y) , , , , , , , , , , ,000,000 12

13 3. K EFF CALCULATIONS Criticality calculations are to be performed for a representative PWR cask model utilizing the PWR spent fuel isotopic compositions from the decay calculations corresponding to the times listed in Table 2. The cask model to be used is described below and is identical to the cask model used in several other of the Expert Group on Burnup Credit Benchmarks. k eff values will be calculated for both actinide only and actinide and fission product cases. The actinide only cases will include 16 O and the nuclides identified as Set1 in Table 1. The actinide and fission product case will include 16 O and the nuclides identified as Set2 in Table 1. Geometry data The representative cask with intact UO assemblies is the criticality model for k eff calculations. The UO 2 assembly geometry and the locations for 25 guide tubes are illustrated in Figure 1. Fuel rod and guide tube radial dimensions are shown in Figures 2 and 3, respectively. Cross-section views of the cask model for use in criticality calculations are provided in Attachment I, Figures 4 through 6. 13

14 Figure 1 : UO 2 assembly geometry and guide tube locations 14

15 Figure 2 : Fuel rod geometry Figure 3 : Guide tube geometry 15

16 Material and geometrical descriptions Fuel assembly Fuel rod data Rod pitch cm Rod length cm Endplug material Zircaloy 4 Endplug height 1.75 cm Full rod length cm (fuel + 2 endplug) Assembly data Lattice (264 fuel rods, 25 guide tubes) Dimensions cm 3 Moderator Water Upper and lower end 50% stainless steel, 50% water (by volume) Hardware (Note: The assembly upper and lower end hardware will be modeled as a region of smeared water and stainless steel; other hardware, such as grid spacers, is ignored). Upper hardware height 30.0 cm Lower hardware height 10.0 cm Upper water region height 7.0 cm Lower water region 0.0 cm Cask Cask shell Inner diameter cm Outer diameter cm Material Stainless steel (SS304) Total height cm Inner cavity height cm Assembly basket Inner basket compartment cm 3 Dimensions Material Borated stainless steel (1 wt% boron) Basket wall thickness 1 cm Configuration 21 assemblies positioned in a 5 5 array (without assembly in corner) Fuel assemblies are centered within basket region Cask is completely flooded with water Material compositions The criticality calculations will use spent fuel compositions from decay calculations (see Section 2) and the nuclide densities for the other assembly and cask materials provided in units of atom/barn cm in this section. 16

17 Spent fuel Spent fuel compositions corresponding to the time cases listed in Table 2 and to the 2 enrichment/burnup pair values (3.0/30 and 4.5/50) will include: 1) actinide-only cases: 16 O and the nuclides identified as Set1 in Table 1. 2) actinides + fission products cases: 16 O and the nuclides identified as Set2 in Table 1. Fuel Clad Fe 1.383E-04 Cr 7.073E-05 O 2.874E-04 Zr 3.956E-02 End plug Cr 7.589E-05 Fe 1.484E-04 Zr 4.298E-02 Guide tube Fe 1.476E-04 Cr 7.549E-05 O 3.067E-04 Zr 4.222E-02 Water Stainless steel H 6.662E-02 O 3.331E-02 Cr 1.743E-02 Mn 1.736E-03 Fe 5.936E-02 Ni 7.721E-03 Borated (1 wt%) Cr 1.691E-02 Stainless steel Mn 1.684E-03 Fe 5.758E-02 Ni 7.489E B 7.836E B 3.181E-03 50/50 stainless steel/ Cr 8.714E-03 Water mixture Mn 8.682E-04 Fe 2.968E-02 Ni 3.860E-03 H 3.338E-02 O 1.669E-02 17

18 4. PARAMETERS REQUIRED 4.1 Fuel compositions Provide compositions for the actinide and fission product isotopes listed in Table 1 (Set 2) for each of the time case numbers listed in Table k eff calculations Provide k eff values for each set of PWR isotopic compositions from the decay calculations for cases involving only the actinides and cases involving both the actinides and the fission products. The total number of k eff calculation cases is 120 (2 enrichment/burnup combinations 30 decay-time steps 2 burnup credit nuclide sets). Criticality calculation cases 1 through 30 will use isotopic compositions for the assembly with 3.0 wt% 235 U enrichment and 30-GWd/MTU burnup that contains actinides only, and correspond to decay time case numbers 1 though 30, respectively. Criticality calculation cases 31 through 60 will use isotopic compositions for the assembly with 3.0 wt% 235 U enrichment and 30-GWd/MTU burnup that contains actinides and fission products, and correspond to decay time case numbers 1 though 30, respectively. Criticality calculation cases 61 through 90 will use isotopic compositions for the assembly with 4.5 wt% 235 U enrichment and 50-GWd/MTU burnup that contains actinides only, and correspond to decay time case numbers 1 though 30, respectively. Criticality calculation cases 91 through 120 will use isotopic compositions for the assembly with 4.5 wt% 235 U enrichment and 50-GWd/MTU burnup that contains actinides and fission products, and correspond to decay time case numbers 1 though 30, respectively. 5. REQUESTED INFORMATION AND RESULTS Forward the results via to the Phase V Benchmark coordinator, John Wagner (wagnerjc@ornl.gov). The results should be provided in three files according to the format instructions provided below. 18

19 5.1 Spent fuel composition results The "spent fuel composition results" will be provided in two files, one file for the PWR assembly and another file for the BWR assembly file. Each file must be composed of: Line No. Contents 1 Type of assembly (PWR or BWR) 2 Date 3 Institute 4 Contact Person 5 address or Telefax Number of the contact person 6 Computer Code initial enrichment and 30-GWd/MTU burnup 8 *Time case 2* 9 Nuclide density (atom/barn cm) of 233 U 10 Nuclide density (atom/barn cm) of 234 U 11 Nuclide density (atom/barn cm) of 235 U 12 Nuclide density (atom/barn cm) of 236 U 13 Nuclide density (atom/barn cm) of 238 U 14 Nuclide density (atom/barn cm) of 238 Pu 15 Nuclide density (atom/barn cm) of 239 Pu 16 Nuclide density (atom/barn cm) of 240 Pu 17 Nuclide density (atom/barn cm) of 241 Pu 18 Nuclide density (atom/barn cm) of 242 Pu 19 Nuclide density (atom/barn cm) of 237 Np 20 Nuclide density (atom/barn cm) of 241 Am 21 Nuclide density (atom/barn cm) of 242m Am 22 Nuclide density (atom/barn cm) of 243 Am 23 Nuclide density (atom/barn cm) of 103 Rh 24 Nuclide density (atom/barn cm) of 133 Cs 25 Nuclide density (atom/barn cm) of 143 Nd 26 Nuclide density (atom/barn cm) of 145 Nd 27 Nuclide density (atom/barn cm) of 155 Gd 28 Nuclide density (atom/barn cm) of 95 Mo 29 Nuclide density (atom/barn cm) of 99 Tc 30 Nuclide density (atom/barn cm) of 101 Ru 31 Nuclide density (atom/barn cm) of 109 Ag 32 Nuclide density (atom/barn cm) of 147 Sm 33 Nuclide density (atom/barn cm) of 149 Sm 34 Nuclide density (atom/barn cm) of 150 Sm 35 Nuclide density (atom/barn cm) of 151 Sm 36 Nuclide density (atom/barn cm) of 152 Sm 37 Nuclide density (atom/barn cm) of 153 Eu 38 *Time case 3* 39 to 67 As for items 9 to * Time case 4* 69 to 97 As for items 9 to * Time case 5* 99 to 127 As for items 9 to * Time case 6* 129 to 157 As for items 9 to * Time case 7* 19

20 159 to 187 As for items 9 to * Time case 8* 189 to 217 As for items 9 to * Time case 9* 219 to 247 As for items 9 to * Time case 10* 249 to 277 As for items 9 to * Time case 11* 279 to 307 As for items 9 to * Time case 12* 309 to 337 As for items 9 to * Time case 13* 339 to 367 As for items 9 to * Time case 14* 369 to 397 As for items 9 to * Time case 15* 399 to 427 As for items 9 to * Time case 16* 429 to 457 As for items 9 to * Time case 17* 459 to 487 As for items 9 to * Time case 18* 489 to 517 As for items 9 to * Time case 19* 519 to 547 As for items 9 to * Time case 20* 549 to 577 As for items 9 to * Time case 21* 579 to 607 As for items 9 to * Time case 22* 609 to 637 As for items 9 to * Time case 23* 639 to 667 As for items 9 to * Time case 24* 669 to 697 As for items 9 to * Time case 25* 699 to 727 As for items 9 to * Time case 26* 729 to 757 As for items 9 to * Time case 27* 759 to 787 As for items 9 to * Time case 28* 789 to 817 As for items 9 to * Time case 28* 819 to 847 As for items 9 to * Time case 30* 849 to 877 As for items 9 to initial enrichment and 50-GWd/MTU burnup 879 *Time case 2* 880 to 908 As for items 9 to 37 20

21 909 *Time case 3* 910 to 938 As for items 9 to * Time case 4* 940 to 968 As for items 9 to * Time case 5* 970 to 998 As for items 9 to * Time case 6* 1000 to 1028 As for items 9 to * Time case 7* 1030 to 1058 As for items 9 to * Time case 8* 1060 to 1088 As for items 9 to * Time case 9* 1090 to 1118 As for items 9 to * Time case 10* 1120 to 1148 As for items 9 to * Time case 11* 1150 to 1178 As for items 9 to * Time case 12* 1180 to 1208 As for items 9 to * Time case 13* 1210 to 1238 As for items 9 to * Time case 14* 1240 to 1268 As for items 9 to * Time case 15* 1270 to 1298 As for items 9 to * Time case 16* 1300 to 1328 As for items 9 to * Time case 17* 1330 to 1358 As for items 9 to * Time case 18* 1360 to 1388 As for items 9 to * Time case 19* 1390 to 1418 As for items 9 to * Time case 20* 1420 to 1448 As for items 9 to * Time case 21* 1450 to 1478 As for items 9 to * Time case 22* 1480 to 1508 As for items 9 to * Time case 23* 1510 to 1538 As for items 9 to * Time case 24* 1540 to 1568 As for items 9 to * Time case 25* 1570 to 1598 As for items 9 to * Time case 26* 1600 to 1628 As for items 9 to * Time case 27* 1630 to 1658 As for items 9 to * Time case 28* 21

22 1660 to 1688 As for items 9 to * Time case 28* 1690 to 1718 As for items 9 to * Time case 30* 1720 to 1748 As for items 9 to Please describe your analysis environment here. It will be included in phase V report. The description should include: Institute and Country, Participants, Neutron data library, Neutron data processing code or method, Description of your code system, Omitted nuclides if any. Omitted cases if any. Other related information. 5.2 k eff values The "keff results" file must be composed of: Line No. Contents 1 keff calculation 2 Date 3 Institute 4 Contact Person 5 address or Telefax Number of the contact person 6 Computer Code 7 3.0/30: actinide only 8 to 37 keff values for cases 1 through 30 (see Section 4.2 for case description) /30: actinides + fission products 39 to 68 keff values for cases 31 through 60 (see Section 4.2 for case description) /50: actinide only 70 to 99 keff values for cases cases 61 through 90 (see Section 4.2 for case description) /50: actinides + fission products 101 to 102 keff values for cases 91 through 120 (see Section 4.2 for case description). 103 Please describe your analysis environment here. It will be included in phase V report. The description should include: Institute and Country, Participants, Description of your code system, Neutron data library, Neutron data processing code or method, Neutron energy groups, Geometry modeling (3-D, 2-D etc.), Omitted nuclides if any. Omitted cases if any. Other related information. 22

23 6. SCHEDULE Assuming this benchmark proposal is finalized and approved by December 2008: June 2009 Participants provide results to Phase V benchmark coordinator September 2009 Distribution of draft Phase V report December 2009 All comments on draft report received by benchmark coordinator April 2010 Final draft of Phase V report for Nuclear Science Committee 23

24 ATTACHMENT I CASK MODEL CROSS-SECTION VIEWS Figures 4 and 5 show top and side views of the cask model. A vertical cross-section through the basket compartment illustrated in Figure 6 shows the fuel rod and assembly geometry regions, including active fuel, rod endplugs, and assembly upper and lower hardware. Figure 4 : Cask model (top view) 24

25 Figure 5 : Cask model (side view) 25

26 Figure 6 : Single basket compartment 26