Study on the prediction accuracy of nuclide generation and depletion with JENDL

Okumura, Keisuke; Oki, Shigeo*; Yamamoto, Munenari*; Matsumoto, Hideki*; Ando, Yoshihira*; Tsujimoto, Kazufumi; Sasahara, Akihiro*; Katakura, Junichi; Matsumura, Tetsuo*; Aoyama, Takafumi*; et al.

JAERI-Research 2004-025, 154 Pages, 2005/01

JAERI-Research-2004-025.pdf:19.46MB

This report summarizes the activity (FY2000-2003) of Working Group (WG) on Evaluation of Nuclide Generation and Depletion under Subcommittee on Nuclear Fuel Cycle of Japanese Nuclear Data Committee. In the WG, analyses of Post Irradiation Examinations have been carried out for UO and MOX fuels irradiated in PWRs, BWRs and FBRs, and for actinide samples irradiated in fast reactors, by using ORIGEN or more detailed calculation codes with their libraries based on JENDL-3.2, JENDL-3.3 and other foreign nuclear data files. From these results, current prediction accuracy and problems for evaluation of nuclide generation and depletion are discussed. Furthermore, this report covers other products of our activity; development of the ORIGEN libraries for PWR, BWR and FBR based on JENDL-3.3, study on introduction of neutron spectrum index to ORIGEN calculations, and results of questionnaire survey on desirable accuracy of ORIGEN calculations.

Analysis of the TRIGA MARK-II benchmark IEU-COMP-THERM-003 with Monte Carlo code MVP

Mahmood, M. S.; Nagaya, Yasunobu; Mori, Takamasa

JAERI-Tech 2004-027, 30 Pages, 2004/03

JAERI-Tech-2004-027.pdf:2.26MB

The benchmark experiments of the TRIGA Mark-II reactor in the ICSBEP handbook have been analyzed with the Monte Carlo code MVP using the cross section libraries based on JENDL-3.3, JENDL-3.2 and ENDF/B-VI.8. MCNP calculations have been also performed with the ENDF/B-VI.6 library for comparison between the MVP and MCNP results. For both cores labeled 132 and 133, which have different core configurations, the ratio of the calculated to the experimental results (C/E) for keff obtained by the MVP code is 0.999 for JENDL-3.3, 1.003 for JENDL-3.2, and 0.998 for ENDF/B-VI.8. For the MCNP code, the C/E values are 0.998 for both the Core 132 and 133. All the calculated results agree with the reference values within the experimental uncertainties. The results obtained by MVP with ENDF/B-VI.8 and MCNP with ENDF/B-VI.6 differ only by 0.02% for Core 132, and by 0.01% for Core 133.

Revised data for 2nd version of Nuclear Criticality Safety Handbook/Data Collection

Okuno, Hiroshi; Ryufuku, Susumu*; Suyama, Kenya; Nomura, Yasushi; Tonoike, Kotaro; Miyoshi, Yoshinori

JAERI-Conf 2003-019, p.116 - 121, 2003/10

https://doi.org/10.11484/jaeri-conf-2003-019

This paper outlines the data prepared for the 2nd version of Data Collection of the Nuclear Criticality Safety Handbook. These data are discussed in the order of its preliminary table of contents. The nuclear characteristic parameters (k, M, D) were derived, and subcriticality judgment graphs were drawn for eleven kinds of fuels which were often encountered in criticality safety evaluation of fuel cycle facilities. For calculation of criticality data, benchmark calculations using the combination of the continuous energy Monte Carlo criticality code MVP and the Japanese Evaluated Nuclear Data Library JENDL-3.2 were made. The calculation errors were evaluated for this combination. The implementation of the experimental results obtained by using NUCEF facilities into the 2nd version of the Data Collection is under discussion. Therefore, related data were just mentioned. A database is being prepared to retrieve revised data easily.

Validation of minor actinide cross sections by studying samples irradiated for 492 days at the dounreay prototype fast reactor, 2; Burnup calculations

Tsujimoto, Kazufumi; Kono, Nobuaki; Shinohara, Nobuo; Sakurai, Takeshi; Nakahara, Yoshinori; Mukaiyama, Takehiko; Raman, S.*

Nuclear Science and Engineering, 144(2), p.129 - 141, 2003/06

https://doi.org/10.13182/NSE03-A2348

To evaluate neutron cross-section data of minor actinides, separated actinide samples and dosimetry samples were irradiated at the Dounreay Prototype Fast Reactor for 492 effective full power days. Based on the burnup calculations of major actinide and dosimetry samples, the neutron flux distribution and the flux level were adjusted at the locations where minor actinide samples were irradiated. The burnup calculations were carried out for minor actinides using the determined flux distribution and flux level. This paper discusses the burnup calculations and the validation of minor actinide cross-section data in evaluated nuclear data libraries. We find that we can obtain reliable FIMA (fission per initial metallic atom) values by using the Nd method except that the uncertainties in the FIMA values are large for U, Pu, Am isotopes, and Cm isotopes because the Nd yields are known poorly for these isotopes and are probably overestimated. For these isotopes, measurements to improve the fission-yield data are needed. We also find that, in general, the JENDL-3.2 nuclear data for the minor actinides are adequate for the conceptual design study of transmutation systems. But, there are some nuclides (especially Pu and Pu) for which new measurements are needed particulary if the minor actinides constitute a major part of the nuclear fuel.

Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

Kuroishi, Takeshi; Hoang, A.; Nomura, Yasushi; Okuno, Hiroshi

JAERI-Tech 2003-021, 60 Pages, 2003/03

JAERI-Tech-2003-021.pdf:4.56MB

The reactivity effect of the asymmetry of axial burnup profile is studied for PWR spent fuel transport cask proposed in OECD/NEA Phase II-C benchmark. The axial burnup profiles are based on in-core flux measurements. Criticality calculations are performed with the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculations are carried out not only for cases in the benchmark but also for symmetric burnup cases. Both actinide-only approach and actinide plus fission product approach is considered. The end effect is more sensitive to higher burnup asymmetry. The axial fission distribution becomes strongly asymmetric as its peak shifts toward the fuel top end. The peak of fission distribution gets higher with the increase of either the burnup asymmetry or the assembly-averaged burnup. The conservatism of uniform axial burnup assumption for the actinide-only approach is estimated quantitatively in comparison with the keff result calculated with experiment-based strongest asymmetric axial burnup profile for the actinide plus fission product approach.

Evaluation of covariances for resolved resonance parameters of U, U, and Pu in JENDL-3.2

Kawano, Toshihiko*; Shibata, Keiichi

JAERI-Research 2003-001, 36 Pages, 2003/02

JAERI-Research-2003-001.pdf:1.67MB

Evaluation of covariances for resolved resonance parameters of , , and was carried out. Although a large number of resolved resonances are observed for major actinides, uncertainties in averaged cross sections are more important than those in resonance parameters in reactor calculations. We developed a simple method which is to provide a covariance matrix for the resolved resonance parameters on the basis of uncertainties in the averaged cross sections. The method was adopted to evaluate the covariance data for some important actinides, and the results were compiled in the JENDL-3.2 covariance file.

Analysis of VENUS-2 MOX core measurements with a Monte Carlo code MVP

Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa

Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 10 Pages, 2002/10

We have analyzed the VENUS-2 MOX core benchmark exercises by using a continuous-energy Monte Carlo code MVP with the nuclear data set JENDL-3.2 and ENDF/B-VI release 5. The VENUS-2 core is cruciform and consists of three fuel regions; the squared central region, the inner and the outer part of the peripheral region are fueled with 3.3% UO, 4.0% UO and MOX. We have constructed 3-D quarter-symmetric calculation model as precisely as possible. All calculations were performed for 200 million histories including 1 million histories of 50 cycles for the initial guess. The C/E values of keff are 1.00500, 0.99793 for JENDL-3.2 and ENDF/B-VI, respectively. They are in good agreement with the experimental one. However, the JENDL-3.2 result overestimates slightly by about 0.5%. For the pin power distribution, the systematic overestimation can be observed in the MOX fuel region. The calculated results tend to underestimate the measured one slightly in the UO fuel regions. However, the dependence on the libraries is not seen.

Adjustment of total delayed neutron yields of U, U and Pu by using results of in-pile measurements of effective delayed neutron fraction

Sakurai, Takeshi; Okajima, Shigeaki

Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 12 Pages, 2002/10

The cross section adjustment method was applied to total delayed neutron yields of U, U and Pu of the JENDL-3.2 file by using experimental results of effective delayed neutron fraction at six cores built in two fast critical facilities of the MASURCA and FCA and a thermal critical facility of the TCA to improve these yields. The adjustment was carried out on the yields given at several incident neutron energy points in the file. Furthermore, to validate these adjusted delayed neutron yields, analyses were performed for the experiments at ZPR fast critical facility. These adjusted yields brought a reduction of uncertainty of calculated and an improvement in agreement of between experiment and calculation.

Analyses of high energy neutron streaming experiments using DUCT-III

Masukawa, Fumihiro; Nakashima, Hiroshi; Sasamoto, Nobuo; Nakano, Hideo*; Tayama, Ryuichi*

Journal of Nuclear Science and Technology, 39(Suppl.2), p.1268 - 1271, 2002/08

http://wwwndc.jaea.go.jp/nd2001/proc/pdf/2_1268.pdf

no abstracts in English

Experiments and analyses on sodium void reactivity worth in uranium-free fast reactor at FCA

Oigawa, Hiroyuki; Iijima, Susumu; Ando, Masaki

Journal of Nuclear Science and Technology, 39(7), p.729 - 735, 2002/07

https://doi.org/10.3327/jnst.39.729

no abstracts in English