TRU production and radiotoxicity Rei Kimura Kenichi Yoshioka Kouji Hiraiwa Shungo Sakurai Satoshi Wada Tsukasa Sugita Toshiba Energy Systems amp Solutions Corporation ID: 1026716
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1. The demand for TRU nuclide cross-sections from the view point of TRU production and radiotoxicity○Rei Kimura, Kenichi Yoshioka, Kouji Hiraiwa, Shungo Sakurai, Satoshi Wada, Tsukasa SugitaToshiba Energy Systems & Solutions Corporation* This study is being performed under the contract with Ministry of Education, Culture, Sports, Science and Technology(MEXT) “MEXT nuclear system research and development Program”.
2. IntroductionObjectiveCalculation Method of the effect of Nuclear DataResults and DiscussionEffect of nuclear data library difference to burn-up compositionDominant energy region of nuclides productionPrioritization of nuclear data library improvement theoretically and experimentally ConclusionContents
3. IntroductionObjectiveCalculation Method of the effect of Nuclear DataResults and DiscussionEffect of nuclear data library difference to burn-up compositionDominant energy region of nuclides productionPrioritization of nuclear data library improvement theoretically and experimentally ConclusionContents
4. Potential radiotoxicity;FP is major factor within 100 yearsTRU (Actinides) is major factor above 100 yearsIntroduction : TRU reduction in LWRJAEA-Data/Code 2010-012, (2010). Time(year)Potential radiotoxicity(Sv/tHM)ActinideActinide(without U, Pu)Fission ProductsActivation ProductsNatural UraniumActinideFPPotential radiotoxicity generated from PWR fuel cycle (Normalized in 1 ton of new fuel)Developing a LWR fuel cycle concept to reduce radiotoxicity
5. Existing plan of nuclear fuel cyclePower generation in Light Water ReactorMOX fuel fabrication in reprocessing plantNuclear transmutation in Fast ReactorIntroduction: Concept of a new fuel cycleLWRFuel PlantReprocessing PlantFRFuel PlantReprocessing PlantSpent FuelUranium-fuelMOX-fuelMOX-fuelStorage FacilityPu, MARadioactive waste generationRadioactive waste reduction
6. A new concept of nuclear fuel cyclePower generation in Light Water ReactorTRU reduction in Light Water ReactorIntroduction: Concept of a new fuel cycleLWRFuel PlantReprocessing PlantFRFuel PlantReprocessing PlantSpent FuelUranium-fuelMOX-fuelMOX-fuelStorage FacilityPu, MARadioactive waste reduction
7. Introduction : Burn-up chain of ActinideResonance absorptionThermal neutron captureMOX fuelUO2 fuelRadiotoxicity composition in the spent fuelMany nuclides are related to the production/annihilation of high-radiotoxic nuclidesHigh-radiotoxicity nuclidesMajor ancestor nuclides of the high-radiotoxicity nuclides
8. PIE benchmark results of JENDLPIE experiment results of Takahama unit-3 [1][1] G. Chiba, et. al, “JENDL-4.0 Benchmarking for Fission Reactor Applications”, J. Nucl. Sci. Technol, 48(2), P172-187 (2011)PIE (Post Irradiation Experiment)results of focused nuclides have large uncertainties or differences from the integral experiment
9. Introduction : Cross section differenceCapture cross section of 238Pu in each librariesThis cross section difference affects to downstream nuclide compositions→ More precise nuclear data are required to more accurate estimation of the nuclide production Difference in librariesLarge uncertainties were guessed
10. ObjectiveExtraction of important nuclides based on the difference of production cross section between libraries and its radiotoxicityExamination of dominant energy region of each important nuclidesRequirements for cross section improvement - nuclides - energy region
11. IntroductionObjectiveCalculation Method of the effect of Nuclear DataResults and DiscussionEffect of nuclear data library difference to burn-up compositionDominant energy region of nuclides productionPrioritization of nuclear data library improvement theoretically and experimentally ConclusionContents
12. Calculation Method of the Effect of Nuclear DataItemConditionCODEMVP-BURNNuclear Data Library※JENDL-4.0u[1](Ref)※ENDF/B-VII.1[2]※JEFF-3.2[3]Burn-up45 GWd/t※libraries for MVP-BURN were used which were provided by JAEABWR fuel bundleEvaluated by typical BWR condition[1] Okumura K.,et.al. "Validation of a Continuous-Energy Monte Carlo Burn-up Code MVP-BURN and Its Application to Analysis of Post Irradiation Experiment", J. Nucl. Sci. Technol., 37, 128 (2000). [2] K. Shibata, et. al. J. Nucl. Sci. Technol. 48(1), 1-30 (2011) [3] M.B. Chadwick, et. al. Nuclear Data Sheet. 112(12) 2887-2996(2011), [5] The JEFF team, http://www.oecd-nea.org/dbdata/jeff, (2014)
13. Calculation Method of the Effect of Nuclear DataU-235U-238Pu-238Pu-239Pu-240Pu-241…Cm-244JENDL-4.0uU-235U-238Pu-238Pu-239Pu-240Pu-241…Cm-244ENDF/B-VII.1，JEFF-3.2Each nuclides effect were evaluated by major nuclides cross section replacing to ENDF or JEFF one by one
14. IntroductionObjectiveCalculation Method of the effect of Nuclear DataResults and DiscussionEffect of nuclear data library difference to burn-up compositionDominant energy region of nuclides productionPrioritization of nuclear data library improvement theoretically and experimentally ConclusionContents
15. IntroductionObjectiveCalculation Method of the effect of Nuclear DataResults and DiscussionEffect of nuclear data library difference to burn-up compositionDominant energy region of nuclides productionPrioritization of nuclear data library improvement theoretically and experimentally ConclusionContents
16. U235-EU238-ENp237-EPu238-EPu239-EPu240-EPu241-EPu242-EAm241-EAm242g-EAm242m-EAm243-ECm242-ECm243-ECm244-EU235-FU238-FNp237-FPu238-FPu239-FPu240-FPu241-FPu242-FAm241-FAm242g-FAm242m-FAm243-FCm242-FCm243-FCm244-FU235U238Nuclide Number density Ratio (ENDF or JEFF / JENDL)Np237Pu238Pu239Pu240Pu241Pu242Am241Am242gAm242mAm243Cm242Cm243Cm244U235U238Np237Pu238Pu239Pu240Pu241Pu242Am241Am242gAm242mAm243Cm242Cm243Cm244Number-density ratio (NRij) = Nij,ENDF or Nij,JEFF /Nij,JENDLRed font：high-radiotoxicity nuclides-9.2%-8.2%-8.2%+3%+4%+4%ENDF/B-VII.1JEFF-3.2Effect of nuclear data library difference NRij results of each nuclide (Nij,ENDF/Nij,JENDL)Changed nuclide (JENDF-4.0u to ENDF/B-VII.1)High radiotoxic nuclides production were affected explicitly in JEFF-3.2 → Particularly, Pu-238, Am-241, Am-243Many TRU nuclides cross sections from JENDL-4.0
17. IntroductionObjectiveCalculation Method of the effect of Nuclear DataResults and DiscussionEffect of nuclear data library difference to burn-up compositionDominant energy region of nuclides productionPrioritization of nuclear data library improvement theoretically and experimentally ConclusionContents
18. Dominant energy region of nuclides productionPu-238Am-241Am-243Differences of reaction rate in specific energies were confirmed↓Realize cross section differences by comparing of librariesEvaluate difference of the energy depend reaction rate as follows“Reaction rate based on theENDF or JEFFーReaction rate based on the JENDL”
19. Difference of reaction rate(ENDF or JEFF - JENDL)Comparing of cross sectionsThe difference of cross section in thermal region (~1.0eV) was confirmed between JEFF and ENDF, JENDLTENDL and latest experiment data also had a differenceCross section difference between libraries –Pu-238-JENDL-4.0ENDF/B-VII.1TENDL-2015JEFF-3.2, EXFOR(A.Chyzh 2013)
20. Elucidation of this difference between experiment and theoretical data by additional experiments and/or theoretical examinations is required for improvement of Pu-238 production estimationComparison latest libraries and experiments data –Pu-238-About 100 barn difference between experiment dataLatest Experiment data in 2013JEFF-3.3 used JENDL-4.0u dataThis is really true value?Incident energy (eV)Cross Section (b)JEFF-3.2JEFF-3.3TENDL
21. Over 1000 barn difference of resonance cross sections between libraries and experiment results were confirmedThese resonance cross sections improvement were required by theoretically and experimentallyCross section difference between libraries –Am-241-Difference of reaction rate(ENDF or JEFF - JENDL)Comparing of cross sections
22. Cross section difference between libraries –Am-243-Difference of resonance cross section at 1.356 eV larger than 1500 b between libraries and/or experiment dataAdditoinally, Std daviatoin around 1000 b↓Improvement of resonance cross section of 1.356eV and 1.744eV is required, theoretically and experimentallyDifference of reaction rate(ENDF or JEFF - JENDL)Comparing of cross sectionsENDF/B-VIIIJENDL-4.0uJEFF-3.3JEFF-3.2TENDL-2015Cross section (b)Incident energy (eV)
23. Cross section difference between libraries –Am-243-Resonance cross section was measured in 2014 by E. Mendoza, et al.Experiment data have large differenceDifference of reaction rate(ENDF or JEFF - JENDL)Comparing of cross sectionsIncident energy (eV)Latest libraries not include this data→ Consideration of latest experiment is required
24. IntroductionObjectiveCalculation Method of the effect of Nuclear DataResults and DiscussionEffect of nuclear data library difference to burn-up compositionDominant energy region of nuclides productionPrioritization of nuclear data library improvement theoretically and experimentally ConclusionContents
25. Prioritization of nuclear data library improvementPrioritynuclidesEnergy rangeReason and comment1Pu-2381meV~1.0eVLarge weight of radiotoxicity in UO2 and MOX spent fuelLatest libraries not considered latest experiment results2Am-243Resonance of 1.356eV and 1.744 eVradiotoxicity of Cm-244 from Am-243 have large impact during 100y from dischargeLatest libraries not considered latest experiment results 3Am-2410.1~100eVLarge impact for long term (Am-241) and short term (Cm-242) radiotoxicity.Large differences were exist between libraries and experiment results
26. Prioritization of nuclear data library improvementPrioritynuclidesEnergy rangeReason and comment1Pu-2381meV~1.0eVLarge weight of radiotoxicity in UO2 and MOX spent fuelLatest libraries not considered latest experiment results2Am-243Resonance of 1.356eV and 1.744 eVradiotoxicity of Cm-244 from Am-243 have large impact during 100y from dischargeLatest libraries not considered latest experiment results 3Am-2410.1~100eVLarge impact for long term (Am-241) and short term (Cm-242) radiotoxicity.Large differences were exist between libraries and experiment resultsSummarized priority of nuclides cross section accuracy improvement from view point of environmental load
27. IntroductionObjectiveCalculation Method of the effect of Nuclear DataResults and DiscussionEffect of nuclear data library difference to burn-up compositionDominant energy region of nuclides productionPrioritization of nuclear data library improvement theoretically and experimentally ConclusionContents
28. Prioritized the requirement for nuclides cross section accuracy improvement1. Pu-238 (1meV~1.0eV)2. Am-243 (Resonance of 1.356eV and 1.744 eV )3. Am-241 ( 0.1 ~ 100eV)Cross section of these nuclides are important to evaluate environmental loadRadiotoxicity, Volume of the final disposal site, etc…Latest libraries of these nuclides used some common dataImportant nuclear data for environmental load estimation were sheared between librariesShould be careful in case of Comparison between nuclear data librariesDifferent libraries have same systematic error and large ( >1000 b ) uncertaintyConclusionWe welcome to make the academia-industry collaboration work for improvement of nuclear data libraries
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30. JENDL-4.0Pu-238:2010,JAEAAm-241:2010,JAEAAm-243:2010,JAEAJEFF-3.2Pu-238:1994,JENDL-3.2Am-241:2012 by CEAAm-243:from JENDL-3.3JEFF-3.3Pu-238:from JENDL-4.0Am-241: same as ENDF/B-VII.1Am-243: from JENDL-4.0ENDF/B-VII.1Pu-238:JENDL-4.0 (~60keV) and 2010,LANLAm-241:resonance is JENDL-4.0, 2010, LANLAm-243:2011, LANL, BNLENDF/B-VIIIPu-238:JENDL-4.0 (~60keV) and 2010,LANLAm-241:resonance is JENDL-4.0Am-243:same as ENDF/B-VII.1Base data of Capture XS, resonance
31. Effect of nuclear data library difference U235-FU238-FNp237-FPu238-FPu239-FPu240-FPu241-FPu242-FAm241-FAm242g-FAm242m-FAm243-FCm242-FCm243-FCm244-F
32. Pu-238Pu-239Radiotoxicity composition in the spent fuelPu-240Pu-241Am-241Cm-242Cm-24431.6%1.0%4.4%23.5%33.3%3.3%2.1%34.3%MOXUO23.4%6.1%36.9%14.7%2.9%1.3%