(PDF) LWR decay heat calculations using a GRS improved ENDF/B-VI based ORIGEN data libraryArticlePDF AvailableLWR decay heat calculations using a GRS improved ENDF/B-VI based ORIGEN data libraryMay 2008DOI:10.1051/ndata:07480Authors: U. HesseU. HesseThis person is not on ResearchGate, or hasn t claimed this research yet. Klemens HummelsheimGesellschaft für Anlagen und Reaktorsicherheit Robert KilgerGesellschaft für Anlagen und Reaktorsicherheit F.E. MoserF.E. MoserThis person is not on ResearchGate, or hasn t claimed this research yet.Show all 5 authorsHide Download full-text PDFRead full-textDownload full-text PDFRead full-textDownload citation Copy link Link copied Read full-text Download citation Copy link Link copiedReferences (17)Figures (3)Abstract and FiguresThe known ORNL ORIGEN code is widely spread over the world for inventory, activity and decay heat tasks and is used stand-alone or implemented in activation, shielding or burn-up systems. More than 1000 isotopes with more than six coupled neutron capture and radioactive decay channels are handled simultaneously by the code. The characteristics of the calculated inventories, e.g., masses, activities, neutron and photon source terms or the decay heat during short or long decay time steps are achieved by summing over all isotopes, characterized in the ORIGEN libraries. An extended nuclear GRS-ORIGENX data library is now developed for practical appliance. The library was checked for activation tasks of structure material isotopes and for actinide and fission product burn-up calculations compared with experiments and standard methods. The paper is directed to the LWR decay heat calculation features of the new library and shows the differences of dynamical and time integrated results of ENDF/B-VI based and elder ENDF/B-V based libraries for decay heat tasks compared to fission burst experiments, ANS curves and some other published data. A multi-group time exponential evaluation is given for the fission burst power of 4 important fission materials, to be used in quick LWR reactor accident decay heat calculation tools. Decay heat of PWR MOX at a burn-up 40 GWd/tHM, calculated with OREST-V04 (ENDF/B-V decay data) compared to OREST-V06 (ENDF/B-VI decay data).…  Library comparison of the decay power and emitted decay energy calculated from figure 7.…  Nuclide numbers and data range LIBMAST06 and former ORIGEN versions.… Figures - uploaded by Robert KilgerAuthor contentAll figure content in this area was uploaded by Robert KilgerContent may be subject to copyright. Discover the world s research20+ million members135+ million publications700k+ research projectsJoin for freePublic Full-text 1Content uploaded by Robert KilgerAuthor contentAll content in this area was uploaded by Robert Kilger on May 04, 2016 Content may be subject to copyright. International Conference on Nuclear Data for Science and Technology 2007DOI: 10.1051/ndata:07480LWR decay heat calculations using a GRS improved ENDF/B-VI basedORIGEN data libraryU. Hessea, Kl. Hummelsheim, R. Kilger, F.E. Moser, and S. LangenbuchGesellschaft f¨ur Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, 85748 Garching, GermanyAbstract. The known ORNL ORIGEN code is widely spread over the world for inventory, activity and decay heattasks and is used stand-alone or implemented in activation, shielding or burn-up systems. More than 1000 isotopeswith more than six coupled neutron capture and radioactive decay channels are handled simultaneously by the code.The characteristics of the calculated inventories, e.g., masses, activities, neutron and photon source terms or the decayheat during short or long decay time steps are achieved by summing over all isotopes, characterized in the ORIGENlibraries. An extended nuclear GRS-ORIGENX data library is now developed for practical appliance. The library waschecked for activation tasks of structure material isotopes and for actinide and fission product burn-up calculationscompared with experiments and standard methods. The paper is directed to the LWR decay heat calculation featuresof the new library and shows the differences of dynamical and time integrated results of ENDF/B-VI based and elderENDF/B-V based libraries for decay heat tasks compared to fission burst experiments, ANS curves and some otherpublished data. A multi-group time exponential evaluation is given for the fission burst power of 4 important fissionmaterials, to be used in quick LWR reactor accident decay heat calculation tools.1 Development of a GRS improved, ENDF/B-VIbased ORIGEN data libraryA new nuclear data library GRS-ORIGENX [11] is now de-veloped for practical appliance. In a first step of development,called LIBMAST04, some problems in the former ORIGENcalculation method [2] and/or in the data libraries for structuralmaterial activation calculations (LIB1), for the actinide build-up (LIB2) and the fission product generation (LIB3) couldbe solved, e.g., the tritium, 14C, 22 Na, 26Al, 60 Fe or 93mNbproduction problem. This was achieved by extending thenumber of neutron reaction channels, the energy groups andthe energy range. All cross sections and build-up channels arecompletely recalculated by point data files JEF-2.2, ENDF/B-VI, JENDL3.2 and EAF97. But the decay data and fissionyields of LIBMAST04 were based on ENDF/B-V as in theburn-up program system OREST-96 [10].In a second step of development LIBMAST06 the decaydata – decay energies, probabilities and channels – and 25fission yield sets are now taken from ENDF/B-VI data bases.The decay energies were analyzed and improved for reactoraccident calculation to avoid the slight under-predictions ofthe reactor decay heat in the first 1000 seconds in ENDF/B-VI. Especially the beta and gamma energies of 70 impor-tant fission products were enlarged by 5%. The library waschecked for structure material isotope activation, and foractinide and fission product burn-up inventories comparedwith experiments and standard calculations. The overall dataaPresenting author, e-mail: Ulrich.Hesse@grs.deof the new library, compared to elder evaluations, are listed intable 1, last column:Table 1. Nuclide numbers and data range LIBMAST06 and formerORIGEN versions.Library ORIGEN ref. [10] LIBMAST-04 LIBMAST-06ref. [2] *) *) **)Number of nuclidesLIB1 253 700 980 1050LIB2 101 144 177 177LIB3 461 847 1116 1195Range of decay dataChannels 78 8 8Fission yield sets 5 5 20 25Delayed n-precursors 0 95 95 225Number of neutron reactionsLIB1 6 6 15 15LIB2 7 7 16 16LIB3 2 2 15 15E-Groups 33 6 6Up to MeV 10 10 20 20*) Decay library based on ENDF/B-V. **) Decay library based onENDF/B-VI.1.1 Burst decay heat exponential presentationsfor 235U, 238 U, 239 Pu and 241PuFor the most important fission materials 235U, 238 U, 239Pu and241Pu we started GRS-ORIGENX decay heat calculations upto 1013 seconds. Due to good agreement in fission burst exper-iments, a time group exponential development was generatedby the GRS-WATT10 code for the four isotopes.©2008 CEA, published by EDP SciencesArticle available at http://nd2007.edpsciences.org or http://dx.doi.org/10.1051/ndata:07480 854 International Conference on Nuclear Data for Science and Technology 2007The data represent the exact ORIGENX results inside one,maximum two percent (see last two lines in the tables) startingat discharge up to 1013 seconds or 317,000 years. They canbe used in quick LWR reactor accident calculations. It shouldbe mentioned, that only the decay heat of the fission productsis respected. Decay heat contributions of actinides and ofcaptured delayed neutrons are excluded.The data fit two simple burst decay power/energy equa-tions, summarizing the time groups m =1.34 for each fissionmaterial l =1.4Pl(t) =mαlm ∗exp(−λlm ∗t),(1)Pl(t) is the power (MeV/s) of material l at time t (s)Ql(t) =m(αlm/λlm )∗1.0−exp(−λlm ∗t),(2)Ql(t) is the emitted energy (MeV) of material l at time t (s).The decay power of fission products for longer burn-uptimes than a fission burst can be constructed by using equation(2) for the reached burn-up time, summing over the fissionrates of the four fission materials and combing this expressionwith equation (1) for the decay time. This method is used, e.g.,in [3].1.2 Burst decay heat calculations compared tomeasurements and other evaluations in theshort time rangeIn figure 1, a comparison of different library evaluations andthe averaged value is made for the burst decay heat of 235U,presented as a curve F*T (power*decay time). Five measure-ments for the total (gamma plus beta) power were used from0.4 up to 7 ×104seconds. It is good to see that LIBMAST06(full square dots) fits the experimental results (empty circles).The ANS curve ([1], triangular dots) slightly over-predicts,but the ENDF/B-V based LIBMAST04 library (empty squaredots) and also the ENDF/B-VI based SCALE5-ORIGENSlibrary (cruciform dots) under-predict the measured data bysome percent.The deviations of codes and single experiments are shownin figure 2, where the zero deviation line is represented by theaveraged five measurements. The spreading of the measure-ments [5–9] against the averaged data is found as ±5%. Themost recent gamma and beta measurements taken from [9](empty square dots, /NZLU35-8/), dated from 1997, are foundup to 5% higher than the zero line, whereas the known ORNL-Dickens data ([6], /NZLU35-2/) are up to 5% lower than theaveraged value.In [9] we found the beta values to be reliable over thewhole time range of measurements, but the gamma valuesafter 2000 seconds are found to be significantly too low, sofor [9] our experimental total results ended at 1000 seconds.Similar results and good agreement in the short time rangewe found for 239Pu in seven measurements from 0.7 to 7×104s, 238U in two measurements from 0.4 to 2 ×104sand 241Pu, where only one measurement from 5 to 104sisavailable.00,20,40,60,811,21,41,61 10 100 1000 10000 100000Decay Time (s)seulaV-T*FLIBMAST06SCALE-5LIBMAST04/ANS-U235/ AVERAGE Fig. 1. Experimental average decay heat 235U in comparison to threelibraries and ANS [1].-15-10-505101520251 10 100 1000 10000 100000Decay Time (s))%( noitaiveDLIBMAST06 SC ALE-5 LIBMAST04/ANS-U235/ /NZLU35-2 /NZLU35-3/NZLU35-7 / /NZLU35-1/ /NZLU35-8/Fig. 2. Deviations of the three libraries and ANS of figure 1 againstexperimental 235U decay heat.We noticed that the pandaemonium effect, the great dis-crepancy between the gamma and beta decay heat in therange between one and 1000 seconds in the ENDF/B-V baseddata libraries ([10] and LIBMAST04) is now avoided in theENDF/B-VI based library of this paper.1.3 Burst decay heat calculations in the longtime rangeThe library was tested against the future German IndustryDecay Heat Standard DIN 24563-2 [3] for PWR MOX fuel.An exponential presentation in 24 time groups is here givenfor four fission materials 235U, 238 U, 239Pu and 241 Pu. Thelong time range of [3] is limited to 109s or ca. 30 years. Inthe whole decay time range a comparison was made with ourLIBMAST06.In the short time range of ref. [3] at burst decay heatcalculations up to 104s we found a slight under-predictionof the averaged 235U experimental data in figure 1 similar toSCALE5/ORIGENS or [6], but for the other fission materials238U, 239 Pu and 241Pu the agreement with our calculation orthe experimental data was very good. For the long time decayheat range greater 104s the DIN-LIBMAST06-comparisonwas made up to 1011s, see figure 3 for 235 U and figure 4 for239Pu. U. Hesse et al.: LWR decay heat calculations using a GRS improved ENDF/B-VI based ORIGEN data library 8550,000010,000100,001000,010000,100001,000001,E+04 1,E+05 1,E+06 1,E+07 1,E+08 1,E+09 1,E+10 1,E+11Decay Time (s)seulaV-T*FU235 D IN 25463 -2 24-gr p.U235 GRS table 2; 34-grp.Fig. 3. 235U decay heat GRS-LIBMAST06 and DIN 25463-2 in thelong time range.0,000010,000100,001000,010000,100001,000001,E+04 1,E+05 1,E+06 1,E+07 1,E+08 1,E+09 1,E+10 1,E+11Decay Time (s)eulaV-T*FPu239 DIN 25463-2 24-grp.Pu239 GRS table 3; 34-grp.Fig. 4. 239Pu decay heat GRS-LIBMAST06 and DIN 25463-2 in thelong time range.The figure 3 for uranium-235, the most important fissionmaterial in UO2fuel, shows clearly that both exponentialpresentations, DIN or LIBMAST06, are in very good agree-ment in a time range up to 1010 s or ca. 317 years. Sameis true for plutonium-239, figure 4, which is in MOX fuelthe most important fission material, and for 238U and 241 Pu(not shown). Beyond such long decay times of 1010 samoredetailed evaluation with more groups than 24 should be used;for example our evaluation, which had been developed up to1013 seconds.2 Burn-up calculations and decay heatFission burst analysis is only one (important) aspect of themethods to prove a library. But in the short pulse all otherneutron activation processes which lead to other long livingdecay heat emitting isotopes are neglected, which normallyoccur in the neutron flux during the reactor burn-up timeperiods. The build-up of 134Cs by neutron capturing in thefission product 133Cs is the most known and most importantexample.Using LIBMAST06 of the GRS-ORIGENX code, six im-proved standard ORIGEN card image libraries were generatedfor our burn-up systems 1D OREST and 3D KENOREST0,0E+005,0E+051,0E+061,5E+062,0E+062,5E+061,E-01 1,E+00 1,E+01 1,E+02 1,E+03 1,E+04 1,E+05 1,E+06 1,E+07 1,E+08Decay Time (s)latemyvaeH noT 1 rep sttaWLIBMAST04 (ENDF/B-V)LIBMAST06 (ENDF/B-VI)Fig. 5. Decay heat of PWR UO2 at a burn-up 40 GWd/tHM,calculated with OREST-V04 (ENDF/B-V decay data) compared toOREST-V06 (ENDF/B-VI decay data).0,940,950,960,970,980,9911,011,021,031,E+00 1,E+01 1,E+02 1,E+03 1,E+04 1,E+05 1,E+06 1,E+07 1,E+08Decay Tim e (s) wen/dlo (Watt) LIBMAST04/LIBMAST06(Joule) LIBMAST04/LIBMAST06Fig. 6. Library comparison of the decay power and emitted decayenergy calculated from figure 5.[4]. In the next figures the heat production of UO2 andMOX PWR fuel after realistic burn-ups of 40 GWd/tHM in astandard fuel assembly is shown for the ENDF/B-VI improvedlibrary LIBMAST06 compared to LIBMAST04, which is theelder ENDF/B-V decay data based library with same updatedneutron cross sections.It should be mentioned, that in the following figuresonly the decay heat of the actinides and fission productsis respected. Decay heat contributions of activated structurematerials and of captured delayed neutrons are excluded.In figure 5 the UO2decay power after the reactor shutdown up to one month is shown for the elder and the improvedlibrary.In figure 6 a quotient old/new is generated for the pro-duction of the decay power and the time integrated emitteddecay energy. The differences of these realistic calculationswith all fission and neutron capturing processes show, that theolder library under-predicts the decay power (full circles) up to1000 s, and under-predicts the emitted energy (empty circles)up to 104s. But it is interesting to see that – starting from thesetime points up to 107s – the elder ENDF/B-V based librarygave a slightly higher decay heat (ca. 1%).Analogous data as shown in figures 5–6 had been gen-erated with the same burn-up and reactor power for a typ-ical MOX fuel assembly as shown in figures 7–8. The 856 International Conference on Nuclear Data for Science and Technology 20071,0E +005,0E +051,0E +061,5E +062,0E +062,5E +061,E+00 1,E+01 1,E+02 1,E+03 1,E+04 1,E+05 1,E+06 1,E+07 1,E+08Decay Time (s) latemyvaeH noT 1 rep sttaWLIBMAST04 (EN DF/B-V)LIBMAST06 (EN DF/B-VI)Fig. 7. Decay heat of PWR MOX at a burn-up 40 GWd/tHM,calculated with OREST-V04 (ENDF/B-V decay data) compared toOREST-V06 (ENDF/B-VI decay data).0,940,950,960,970,980,9911,011,021,031,E+00 1,E+01 1,E+02 1,E+03 1,E+04 1,E+05 1,E+06 1,E+07 1,E+08Decay Time (s) wen/dlo (Watt) LIBMAST04/LIBMAST06(Joule) LIBMAST04/LIBMAST06Fig. 8. Library comparison of the decay power and emitted decayenergy calculated from figure 7.maximum MOX decay heat under-predictions arrive 5%, theover-prediction in the time range 104–108s arrive a maximumof 2%.3 ConclusionsA new GRS-ORIGENX library was developed with 16 neu-tron reactions and 8 decay channels based on ENDFB/F-VIdecay data and fission yields and cross section point datafor 500 capturing isotopes. The library is used in activation,shielding and burn-up tasks. In the short time range thedecay energies of the 70 most important ENDF/B-VI fissionproduct isotopes were enlarged by 5%. For a comparisongood agreement with experimental fission burst decay heatexperiments for 235U, 238 U, 239Pu and 241 Pu could be shown.A multi-group time exponential evaluation was generated,which can be used in quick reactor accident calculations,representing the exact results of GRS-ORIGENX inside 1% ormaximal 2% from discharge up to 1013seconds. A long timecalculation compared to the future German Industry DecayHeat Standard for PWR MOX fuel is in good agreementbetween 1up to 1010 seconds for 238 U, 239Pu and 241 Pu andbetween 104up to 1010 seconds for 235U. The library isavailable in the GRS-ORIGENX format or in the standardORIGEN format for the HTGR, LWR and FBR reactor type.References1. A. Tobias, Berkeley Nuclear Laboratories, Uncertainties in JEF1Integral Decay Heat Predictions,Proceedings of a Specialists’Meeting on Data for Decay Heat Predictions, Studsvik, Sweden,7–10 Sept. 1987, table 9, p. 243.2. M.J. Bell, ORIGEN - The ORNL Isotope Generation and Deple-tion Code, ORNL-4628, UC-32-Mathematics and Computers,May 1973.3. Berechnung der Zerfallsleistung der KernbrennstoffevonLeichtwasserreaktoren – Teil 2: Uran-Plutonium-Mischoxid(MOX)-Kernbrennstofff¨ur Druckwasserreaktoren, DeutscheNorm – Entwurf, DIN-25463-2, 2004.4. U. Hesse, T. Voggenberger, F. Cester, R. Kilger, KENOREST-2004, Eingabebeschreibung f¨ur quadratische Gitter;GRS5,Sept. 2005.5. M.F. James, Libraries of Evaluated Fission Yields, Evaluation ofFission Product Yields,Proceedings of a Specialists’ Meeting onData for Decay Heat Predictions, Studsvik, Sweden, 7–10 Sept.1987, table 13, p. 101.6. J.K. Dickens, Oak Ridge National Laboratory, Tennessee 37831USA, Review of New Integral Determinations of Decay Heat,same meeting, ORNL-Data, figure 2, p. 204.7. J.K. Dickens, Oak Ridge National Laboratory, Tennessee 37831USA, same meeting, NERL-Data, figure 2, p. 204.8. P.I. Johannson, University of Uppsala, Nykoeping, Sweden,Integral Determinations of the Beta and Gamma Heat,samemeeting, table 1, p. 217, table 2, p. 218.9. H.V. Nguyen et al., Decay Heat Measurements Following Neu-tron Fission of 235-U and 239-Pu,Int. Conf. on Nuclear Datafor Science and Technology, Trieste, Italy, 1997.10. U. Hesse, J. Sieberer; OREST-96, User-Instruction, GRS 6, July1999.11. U. Hesse, E. Moser, Kl. Hummelsheim, GRS-ORIGENX – 2004,Eingabebeschreibung, GRS 6, Sept. 2005.Citations (0)References (17)ResearchGate has not been able to resolve any citations for this publication.Gamma-Ray Decay Heat Measurements Following Neutron Fission of ^235U.^ArticleOct 1996H. V. NguyenJ. M. CampbellG. P. CouchellT. R. EnglandA 5 x5 NaI(Tl) spectrometer was used in the measurement of aggregate gamma-ray energy spectra of fission products resulting from the thermal neutron fission of ^235U. The measurements cover an energy range of 0.1-8.0 MeV and a delay time range of approximately 0.1-14000s. A helium-jet/tape transport system was used in this study which facilitated the measurements at short delay times. In obtaining the energy distribution, the measured spectra were corrected for the response of the spectrometer to monoenergetic gamma-rays. From these energy distributions, the relative gamma decay heat as a function of delay time has been determined. A comparison of the measured gamma decay heat with summation calculations using the code CINDER10 and the ENDF/B-VI database will be presented. Supported in part by the U.S. Department of EnergyViewShow abstractORIGEN: the ORNL isotope generation and depletion codeArticleMay 1973M.J. BellViewUncertainties in JEF1 Integral Decay Heat Predictions, Proceedings of a Specialists Meeting on Data for Decay Heat PredictionsOct 19877-10A TobiasA. Tobias, Berkeley Nuclear Laboratories, Uncertainties in JEF1 Integral Decay Heat Predictions, Proceedings of a Specialists Meeting on Data for Decay Heat Predictions, Studsvik, Sweden, 7–10 Sept. 1987, table 9, p. 243.ORIGEN -The ORNL Isotope Generation and Depletion Code, ORNL-4628, UC-32-Mathematics and ComputersApr 1973M J BellM.J. Bell, ORIGEN -The ORNL Isotope Generation and Depletion Code, ORNL-4628, UC-32-Mathematics and Computers, May 1973.Oct 2004U HesseT VoggenbergerF CesterR KilgerU. Hesse, T. Voggenberger, F. Cester, R. Kilger, KENOREST-2004, Eingabebeschreibung für quadratische Gitter; GRS 5, Sept. 2005.Libraries of Evaluated Fission Yields, Evaluation of Fission Product Yields, Proceedings of a Specialists Meeting on Data for Decay Heat PredictionsOct 19877-10M F JamesM.F. James, Libraries of Evaluated Fission Yields, Evaluation of Fission Product Yields, Proceedings of a Specialists Meeting on Data for Decay Heat Predictions, Studsvik, Sweden, 7–10 Sept. 1987, table 13, p. 101.Tennessee 37831 USA, Review of New Integral Determinations of Decay Heat, same meeting, ORNL-Data204J K DickensJ.K. Dickens, Oak Ridge National Laboratory, Tennessee 37831 USA, Review of New Integral Determinations of Decay Heat, same meeting, ORNL-Data, figure 2, p. 204.Integral Determinations of the Beta and Gamma Heat, same meeting218P I JohannsonP.I. Johannson, University of Uppsala, Nykoeping, Sweden, Integral Determinations of the Beta and Gamma Heat, same meeting, table 1, p. 217, table 2, p. 218.OREST-96, User-InstructionAug 1999U HesseJ SiebererU. Hesse, J. Sieberer; OREST-96, User-Instruction, GRS 6, July 1999.Uncertainties in JEF1 Integral Decay Heat PredictionsSep 1987243A TobiasA. Tobias, Berkeley Nuclear Laboratories, Uncertainties in JEF1 Integral Decay Heat Predictions, Proceedings of a Specialists Meeting on Data for Decay Heat Predictions, Studsvik, Sweden, 7-10 Sept. 1987, table 9, p. 243.Show moreAdvertisementRecommended publicationsDiscover moreArticleFull-text availableDevelopment of modern CANDU PHWR cross-section libraries for SCALEJune 2016 · Nuclear Engineering and DesignNathan T. Shoman Steven Eugene SkutnikA new set of SCALE fuel lattice models have been developed for the 28-element and 37-element CANDU fuel assembly designs using modern cross-section data from ENDF-B/VII.0 in order to produce new reactor data libraries for SCALE/ORIGEN depletion analyses. These new libraries are intended to provide users with a convenient means of evaluating depletion of CANDU fuel assemblies using ORIGEN through ... [Show full abstract] pre-generated cross sections based on SCALE lattice physics calculations. The performance of the new CANDU ORIGEN libraries in depletion analysis benchmarks to radiochemical assay data were compared to the previous version of the CANDU libraries provided with SCALE (based on WIMS-AECL models).Benchmark comparisons with available radiochemical assay data indicate that the new cross-section libraries perform well at matching major actinide species (U/Pu), which are generally within 1–4% of experimental values. The library also showed similar or better results over the WIMS-AECL library regarding fission product species and minor actinoids (Np, Am, and Cm). However, a notable exception was in calculated inventories of ¹⁵⁴Eu and ¹⁵⁵Eu, where the new library employing modern nuclear data (ENDF/B-VII.0) performed substantially poorer than the previous WIMS-AECL library (which used ENDF-B/VI.8 cross-sections for these species). The cause for this discrepancy appears to be due to differences in the ¹⁵⁴Eu thermal capture cross-section between ENDF/B-VI.8 and ENDF/B-VII.0, an effect which is exacerbated by the highly thermalized flux of a CANDU heavy water reactor compared to that of a typical light-water reactor spectrum.Overall, the new SCALE-based CANDU libraries appear to give very good agreement to experimental benchmarks, indicating that the new libraries should provide satisfactory depletion calculation performance generally comparable or better to the prior WIMS-AECL libraries.View full-textArticleANALYSIS OF LWR BENCHMARKS BASED ON DIFFERENT METHODS AND NUCLEAR DATA EVALUATIONSS. Langenbuch Winfried Zwermann Wolfgang Bernnat[...]M. MattesFor reactor physics applications, cross section data sets based on the latest versions of the available evaluated nuclear data files JEF-2.2, JEFF-3, ENDF/B-VI up to release 5, and JENDL-3.2 were generated for continuous Monte Carlo and multigroup transport calculations. The libraries contain nuclides for actinides, structure materials, and fission products, necessary for LWR problems with fresh ... [Show full abstract] or irradiated fuel for the temperature range from 293 K up to 3000 K. They were prepared with NJOY version 97 using a reconstruction accuracy of 0.1 % for all data sets. The data libraries mainly will be used for calculating safety related reactor physics parameters for standard and innovative fuel in core configurations under operational and accidental conditions, and criticality safety as well. As transport codes, the continuous Monte Carlo code MCNP and multigroup codes based on S N (e.g. XSDRN) or Monte Carlo (e.g. KENO-Va or KENO-VI) methods will be used. The calculation of weighted cross sections for the multigroup codes were performed by the first collision spectral code RESMOD solving the slowing down equation for 26,000 energy points. Since today several data evaluations are available, for main nuclides data sets were generated for different evaluations to see their influence on integral reactor parameters. To verify the generated libraries, numerous benchmark experiments relevant for LWR systems were analysed and compared with measured values. The conclusion of these comparisons is that for homogeneous systems, generally JENDL-3.2 based calculations give higher k eff values than JEF-2.2 or ENDF/B-VI based values. A number of heterogeneous UOX and MOX systems agree very well with experiments if JENDL-3.2 based data were used, JEF-2.2 and ENDF/B-VI based calculations lie always at the lower bound of the given experimental uncertainty. Calculations for heterogeneous UOX systems based on the new release of data for U-235 and U-238 for ENDF/B-VI (Release 5) generally show too low k eff 2 values (up to 1 %). For some homogeneous highly enriched uranium nitrate water solutions as well as for plutonium nitrate water solutions, the JENDL based calculations overpredict the measured values, while the JEF-2.2 and ENDF/B-VI based calculations agree very well with experimental data.Read moreArticleIsotopic inventory calculations taking into account 2D/3D environment conditions during fuel irradia...January 2008 Robert KilgerU. HesseS. LangenbuchBurnup-dependent 2D or 3D flux calculation tools coupled to rod-wisely applied point depletion codes being used to model a whole fuel assembly nowadays are state of the art in inventory calculation during in-core fuel depletion. By these tools, radial structures in modern fuel assembly designs can be accounted for properly in models were infinite assembly arrays are an acceptable approximation. ... [Show full abstract] However, in some cases this infinite array assumption is strongly violaled, e.g. when applied to a MOX fuel assembly surrounded by uranium dioxide assemblies, or when dealing with an assembly being located at the core periphery for one or more irradiation cycles. Moreover, the inventory calculation for spent nuclear fuel samples taken from rod positions near to the very top or bottom of a fuel assembly suffers from neutron leakage and backscatter effects from the environment and thus spectral shifts which cannot be regarded by two dimensional flux models. This paper demonstrates the improved capabilities of the GRS 2D/3D depletion calculation system KENOREST in dealing with the above mentioned difficulties on the basis of inventory calculations for the PIE samples ARIANE BM5 and Takahama-3 SF97-1.Read moreConference PaperFull-text availableNeutron activation of reactor components during operation lifetime of a NPPOctober 2007U. Hesse Gunter Guido PretzschB. Gmal Klemens HummelsheimIntroductionThe knowledge of the level of activation of materials, which have been exposed to high neutron irradiation during lifetime of a nuclear facility, is important for decommissioning and for lifetime extension as well if this is intended. Besides direct measurement of material probes, the calculation of material activation can provide useful and important information with respect to the ... [Show full abstract] long term irradiation behavior of the material of interest. The presentation gives an overview on state of art calculation methods for activation and shows examples of application with respect to decommissioning of NPP.State of the artSince several years GRS uses own-developed calculation tools for material activation, where the well known ORNL code ORIGEN is applied as a main tool. The standard method is the GRSAKTIV-II code system, where ORIGEN runs in a loop over multiple material regions with different irradiation conditions of neutron flux strength and spectra, but with the same irradiation time history. Precalculated multigroup neutron fluxes and cross sections are used in 84 neutron groups. The ORIGEN libraries inside GRSAKTIV-II are based on ENDF/B-V with 6 nuclear reaction channels and 3 neutron energy groups up to 10 MeV. Development of advanced methods and libraries for activation calculationsCurrently an extended version GRS-ORIGENX including new updated libraries based on modern nuclear point data files is being developed for practical application with 15 nuclear reaction channels and 6 neutron energy groups up to 20 MeV. In former versions of ORIGEN only 10 irradiation time steps could be used. Now a maximum of 999 time steps can be handled in double precision mode. Some problems existing in the former ORIGEN calculation method and in its data libraries concerning structural material activation calculations, the actinide build-up and fission product generation can now be solved, e.g. the Tritium, Na-22, Al-26, Fe-60 or Nb-93m generating problem. The decay data are taken from ENDF/B-VI data bases. Due to known contaminations of structure materials by uranium and thorium, the build-up and depletion chains of the heavy metal isotopes can also completely be recalculated in the same way as the build-up chains of induced fission products. More than 20 fission yield sets are taken from ENDF/B-VI data bases. The new generated ORIGEN libraries have also been successfully checked for reactor decay heat conditions. The GRS AAA_Activation SequenceExecuting activation calculations in the environment of a nuclear reactor one has to solve three parts of calculations, what is called a full AAA_sequence. The abbreviation AAA stands for the German words Abbrand as burn-up of fuel, Abschirmung as attenuation of neutrons and gammas and Aktivierung as activation of the irradiated materials. Firstly 1d/2d/3d burn-up calculations KENO/OREST have to be started to find the neutron flux strength in the half-burned reactor core region. Secondly more dimensional multigroup deep-penetration transport calculations DORT from the core region to the chosen structure material region must be done to find the attenuation factors and the neutron spectra. The neutron spectra are necessary to achieve here the correct problem dependent neutron cross sections. And lastly the activation calculations have to be done by GRS-ORIGENX for the structural material regions. For each part of the AAA_sequence a consistent set of burn-up, transport and activation libraries will be used: All neutron cross sections will be completely recalculated by point data files JEF2.2, ENDF/B-VI, JENDL3.2 and EAF97 for 500 structure material isotopes, heavy metal isotopes and fission product isotopes. Although in different data formats of the different codes, the same neutron cross sections will be used in the burn-up, the transport and the activation step of the AAA_sequence. The two first parts of the AAA_sequence have now already been programmed in a closed GRS code system called DORTABLE for practical appliance, which consists of the two main code systems OREST and DORT and some interface tools for data transfer and cross section handling, running in a UNIX or LINUX environment. It is planned in the future to include additionally the extended GRS-ORIGENX code, which at moment is being started by the user as a stand-alone version. Comparison of calculations with experimentsAt first we will show in this paper the application of the new AAA_sequence for activation calculations of the steel upper and lower parts of an UO2 assembly, which was irradiated four cycles in a PWR. The results are compared with measurements, published 1987. At second we will use the AAA_sequence for the irradiation of a test material of the biological shield, irradiated for one hour in a Japanese research reactor to analyze the impurities in concrete. The results will be compared with published measurements.Applications with regard to the operation lifetime of a real NPP In this paper new results of the AAA_sequence for the operation lifetime of a real NPP will be presented. The sequence will start in a simplified whole core model generating realistic neutron flux data for a PWR type reactor. The average burn-up, the flux strength during burn-up, the inventories and the cross sections will be calculated by OREST in 83 energy groups. In the second part of the sequence the neutron flux data will be achieved by 2d deep penetration neutron transport calculations using the SN code DORT in 83 or in 175 energy groups. The up-scatter procedure will be done in 32 groups to achieve the correct flux shape in the thermal energy range of the neutrons. The anisotropy of the scattering will be handled by the PL order of 3 up to 5. The meshes of the 2d DORT core/vessel/shield model will be automatically calculated by the mesh generator, included in the DORTABLE system. Activation of the NPP vessel The vessel has been divided into five layers which are activated separately by GRS-ORIGENX. The variations of the neutron flux strength, the spectra and most important build-up cross sections inside the vessel, calculated by DORT and interface tools, are shown at different geometric points. Evaluation of the results with respect to nuclides of interest, e.g. from the point of view of radiation protection and waste management, will be presented. A comparison with results of earlier calculations will also be presented. It will be shown that back-scattering effects of neutrons from the adjacent biological shield are important for the activation of the adjacent vessel layer.Activation of the NPP biological shieldThe biological shield has been divided into five layers for which the irradiation has been calculated separately by GRS-ORIGENX. The variations of the neutron flux strength, the spectra and most important build-up cross sections inside the concrete different geometric points will be presented. Evaluation of the results with respect to nuclides of interest, e.g. from the point of view of radiation protection and waste management, will be presented. A comparison with results of earlier calculations will also be presented.Analyses of important radioactive isotopes in the short, intermediate and long term Our analyses consist of three parts: the short term region up to 1 year after shut-down, the intermediate term region up to 100 years for intermediate storage of the components of the reactor, and the long term region up to 10.000 years responsible for final storage considerations. A maximum life time of 40 years of reactor operation has been assumed. Additionally, the influence of reactor life times reaching 50 years has been investigated.View full-textArticleNuclear Codes and Data Libraries at GRS for Reactivity and Nuclide Inventory CalculationsU. HesseS. Langenbuch Winfried ZwermannAn overview is given on the nuclear codes and data libraries that are applied at GRS for reactivity and nuclide inventory calculations. The programme system KENOREST was developed for reactivity and burnup calculations by coupling the 3d Monte Carlo code KENO with the 1d burnup code OREST. The capability of this code system will be presented together with results for typical applications. In ... [Show full abstract] addition, results of calculations for critical experiments will be presented that have been performed with Monte Carlo codes MCNP and KENO as well as with deterministic neutron transport codes for the qualification of these codes and the nuclear data libraries generated on the basis of JEF-2.2.Read moreDiscover the world s researchJoin ResearchGate to find the people and research you need to help your work.Join for free ResearchGate iOS AppGet it from the App Store now.InstallKeep up with your stats and moreAccess scientific knowledge from anywhere orDiscover by subject areaRecruit researchersJoin for freeLoginEmail Tip: Most researchers use their institutional email address as their ResearchGate loginPasswordForgot password? Keep me logged inLog inorContinue with GoogleWelcome back! Please log in.Email · HintTip: Most researchers use their institutional email address as their ResearchGate loginPasswordForgot password? 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