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Tada, Kenichi; Endo, Tomohiro*
EPJ Web of Conferences, 284, p.14013_1 - 14013_4, 2023/05
Times Cited Count：0 Percentile：0.21(Nuclear Science & Technology)The self-shielding effect in the unresolved resonance region has a large impact on the fast- and intermediate-spectrum reactors. The probability table method is widely used for continuous-energy Monte Carlo calculation codes to treat the effect. In this method, a table provides the probability distribution of the cross-section for a nuclide in the given energy grid points. The table is generated by averaging with a lot of "ladders" which represent pseudo resonance structures. Though many nuclear data processing codes require the number of ladders as an input parameter to generate the probability table, an optimal number of ladders has not been investigated. Our previous study revealed that the suitable number of ladders depends on the nuclide and its resonance parameters. This result indicates that it is very difficult for users to find the optimal number of ladders. We developed the calculation method of the statistical uncertainty for the probability table generation.
Ueki, Taro
Nuclear Science and Engineering, 193(7), p.776 - 789, 2019/07
Times Cited Count：5 Percentile：48.18(Nuclear Science & Technology)It is known that the convergence of standardized time series (STS) to Brownian bridge yields standard deviation estimators of the sample mean of correlated Monte Carlo tallies. In this work, a difference scheme based on a stochastic differential equation is applied to STS in order to obtain a new functional statistic (NFS) that converges to Brownian motion (BM). As a result, statistical error estimation improves twofold. First, the application of orthonormal weighting to NFS yields a new set of asymptotically unbiased standard deviation estimators of sample mean. It is not necessary to store tallies once the updating of estimator computation is finished at each generation. Second, it becomes possible to assess the convergence of sample mean in an assumption-free manner by way of the comparison of power spectra of NFS and BM. The methodology is demonstrated for three different types of problems encountered in Monte Carlo criticality calculation.
Usami, Shin; Kishimoto, Yasufumi*; Taninaka, Hiroshi; Maeda, Shigetaka
JAEA-Technology 2018-003, 97 Pages, 2018/07
JAEA-Technology-2018-003.pdf:12.54MB
The decay heat used for effectiveness evaluation of the prevention measures against severe accidents in the prototype fast breeder reactor Monju was evaluated by applying the updated nuclear data libraries based on JENDL-4.0, reflecting the realistic core operation pattern, and setting the rational extent of uncertainty. The decay heats of fission products, the actinide nuclides such as Cm-242, and radioactive structural materials were calculated by FPGS code. The decay heat of U-239 and Np-239 was evaluated based on ANSI/ANS-5.1-1994. The calculation uncertainty of each decay heat was evaluated based on summation of uncertainty factors, C/E values of reaction rates obtained in Monju system startup test, and so on. Furthermore, the decay heat evaluation method based on the FPGS90 was verified by the comparison of the results of the decay heat measurement of the two spent MOX fuel subassemblies in the experimental fast reactor Joyo MK-II core.
Matsunaga, Go; Takechi, Manabu; Sakurai, Shinji; Suzuki, Yasuhiro*; Ide, Shunsuke; Urano, Hajime
Fusion Engineering and Design, 98-99, p.1113 - 1117, 2015/10
Times Cited Count：15 Percentile：77.35(Nuclear Science & Technology)Miyamoto, Yutaka
Hoshaka Bunseki Handobukku, 7 Pages, 2004/00
no abstracts in English
Nuclear Code Evaluation Special Committee of Nuclear Code Research Committee
JAERI-Review 2002-003, 97 Pages, 2002/03
JAERI-Review-2002-003.pdf:8.64MB
no abstracts in English
2000/12)Kobayashi, Hideo; ; ; ; ; *; *
JNC TN8440 2001-005, 33 Pages, 2001/02
JNC-TN8440-2001-005.pdf:1.24MB
Four mass spectrometers are used for plutonium and uranium isotopic composition analysis in the Plutonium Fuel Center. Analytical errors of the analysis was evaluated based on routine control data obtained with measurement of reference materials. It was confirmed that the errors satisfied the International Target Values for safeguard analysis, and the random error of the analysis was improved by changing conventional method to total evaporation method.
; Sato, Wakaei*; Iwai, Takehiko*
JNC TN9400 2000-096, 113 Pages, 2000/06
This report describes the updated analyses results on the BFS-58-1-I1 core. The experiment was conducted at BFS-2 of Russian Institute of Physics & Power Engineering (IPPE). The central region is "non-Uranium fuel zone", where only Pu can induce fission reaction. The non-U zone is surrounded by MOx fuel zone, which is surrounded by U0
fuel zone. Sodium is used for simulating the coolant material. As it was found that the lattice pitch had been incorrectly understood in the past analyses, all items have been re-calculated using the corrected number densities. Furthermore, significantly softened neutron spectrum in the central region caused problems in applying the plate-stretch model that has been established for fast reactor cores through JUPITER experimental analyses. Both keeping the pellet density and using SRAC library for the elastic cross section for lighter nuclides allow us to obtain reasonable analysis accuracy on the spectral indices that were measured at the center of the core. Application of such a cell model was justified through comparison among various cell models using continuous energy Monte-Carlo code MVP. It is confirmed that both the MOX zone and the U0 zone can be correctly evaluated by the plate-stretch model. Based on the updated cell calculation, both the effective multiplication factor (k-eff)and the spectral indexes agree well with the measured values. The transport and mesh-size correction is made for the k-eff evaluation. Those results also agree well within reasonable difference between those obtained by IPPE and CEA, which were obtained by using sub-group method or continuous-energy Monte Carlo code. Evaluation by the nuclear data library adjustment confirmed that the analyses results of the BFS-58-1-I1 core have no significant inconsistency with JUPITER experimental analyses results. Those results are quite important for starting BFS-62 cores, which will be analyzed in the framework of supporting program for Russian ...
; ; ; Ikeda, Hisashi ; Jitsukata, Shu*; *
JNC TN8410 2000-022, 55 Pages, 2000/05
JNC-TN8410-2000-022.pdf:1.57MB
Measurement of U and Pu concentrations by wavelength dispersion type X-ray fluorescence spectrometry was studied. Sample holder was installed inside of glove box and other instruments, X-ray tube, monochromator and detectors were set out side of the glove box. X-rays was irradiated to sample though Be window. Fluorescent X-rays form sample were also passing though the same Be window and detected outside. Analytical conditions were optimized as follows. Sample thickness is 8 mm, which is 3ml of sample volume by the sample holder. Voltage and eurrent for X-ray tube is 50kV and 40 mA, respectively. Measurement was done twice, 60 seconds each, and averaged X-ray intensity was used to calculate elemental concentrations. Matrix correction was necessary to measure U and Pu concentration within 10% accuracy. Detection limits were calculated to 0.4 mg/L for U and 0.7mg/L for Pu. Calibration curve was liner up to 9 g/L fbr U and Pu. Two calculation methods, calibration curve method and standard addition method, were studied to measure Pu concentration in organic solution. Detection limit was 5.3 mg/L and 0.2 mg/L, respectively.
; Numata, Kazuyuki*
JNC TN9400 2000-036, 138 Pages, 2000/03
JNC-TN9400-2000-036.pdf:10.16MB
Japan Nuclear Cycle Development lnstitute (JNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were renected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. ln this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor "JOYO" MK-l core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. 0n the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of "JOYO" MK-l core in comparison with ZPPR-9 core of JUPITER experiments.
*; Kitada, Takanori*; Tagawa, Akihiro; *; Takeda, Toshikazu*
JNC TJ9400 2000-006, 272 Pages, 2000/02
JNC-TJ9400-2000-006.pdf:9.69MB
Investigation was made on the follwing three themes as a part of the improvement of reactor physics analysis method for FBR with various core concept. Part 1: Investigation of Error Estimation of Neutron Spectra in FBR and Suggestions to Improve the Accuracy. In order to improve the spectrum unfolding method used in fast experimental reactor JOYO, a trial was made to evaluate the error in the estimated neutron spectrum, cause by cause. And the evaluated errors were summed up to obtain the most probable and reasonable error as possible. The summed up error was found relatively small compared to the error caused by the uncertainty of cross section data: most of the error in the spectrum unfolding method can be attributed to the error in cross sections. It was also found that the error due to the fission spectrum causes a considerable error in the high energy neutron spectrum which is over several MeV. Part 2: Study on Reactor Physics Analysis Method for Gas-Cooled FBR. In gas-cooled FBR, the portion of coolant channels in core volume is larger than sodium-cooled FBR. This leads to strong neutron streaming effects. For sodium-cooled FBR, several methods were proposed to evaluate the neutron streaming effect, however, these methods can not be used directly to gas-cooled reactor because the direction dependent diffusion coefficient becomes infinitive along the direction pararel to the coolant chammel. In this study, a new method is proposed to evaluate the neutron streaming effect, based on the method taking the axial buckling into consideration, which method was originally proposed by K
hler. Part 3: Study on Reactor Physics Analysis Method for Water-Cooled FBR An investigation was made on low-moderated water-cooled FBR, on the point that the ordinary used analysis method for FBR may give considerable difference in results in such core. In light water reactors, it is well known that the space dependence of self-shielding effect of heavy nuclides are considerably ...
*; *
JNC TJ9400 2000-004, 109 Pages, 2000/02
JNC-TJ9400-2000-004.pdf:4.96MB
We estimated covariances of the JENDL-3.2 data on the nuclides and reactions needed to analyze fast-reactor cores for the past three years, and produced covariance files. The present work was undertaken to re-examine the covariance files and to make some improvements. The covariances improved are the ones for the inelastic scattering cross section of 
O, the total cross section of 
Na, the fission cross section of 
U, the capture cross section of 
U, and the resolved resonance parameters for U. Moreover, the covariances of 
U data were newly estimated by the present work. The covariances obtained were compiled in the ENDF-6 format.
Shibamoto, Yasuteru; Kukita, Yutaka*; Nakamura, Hideo; Park, H. S.*; Anoda, Yoshinari
Proceedings of 8th International Conference on Nuclear Engineering (ICONE-8) (CD-ROM), p.12 - 0, 2000/00
no abstracts in English
Naito, Osamu; Yoshida, Hidetoshi; Kitamura, Shigeru; *; *
Review of Scientific Instruments, 70(9), p.3780 - 3781, 1999/09
Times Cited Count：17 Percentile：71.61(Instruments & Instrumentation)no abstracts in English
Sasamoto, Hiroshi; Yui, Mikazu; Savage, D.*; Bille, B.*
JNC TN8400 99-025, 32 Pages, 1999/06
Groundwater data used for modelling site or repository evolution need to be assessed for their quality and whether they are "fit for purpose", prior to utilization. This report discuss factors and issues which impinge upon the quality of such data. It is recommended that geochemical modelleres : (1)are aware of how groundwater samples were collected, whether during drilling, during hydraulic testing, or thereafter, by in-situ measurement, pumped from boreholes, or by pressurised sampler ; (2)are aware of what procedures (if any) were used to "correct" samples for drill fluid contamination and what errors were associated with those methods ; (3)are aware of whether samples were subject to de-pressurisation during sampling, and whether geochemical modelling techniques were applied to correct the compositions of samples for that process ; (4)request different measures of redox activity (e.g., electrode measurements of Eh, concentrations of different redox-sensitive aqueous species) to be applied to key groundwater samples to investigate the extent of redox equilibrium ; (5)are aware of how groundwater samples were filtered and preserved for off-site analysis ; (6)ensure that adequate methods of groundwater filtration (
0.1
m) and chemical analysis are applied to ensure accurate and reproducible analyses for dissolved aluminum at low levels of concentration (generally less than 0.2 mg/L) ; (7)are aware of elemental errors and detection limits in chemical analysis of groundwater samples and assess the quality of groundwater analyses via ion exchange balances and via a comparison of measured and calculated values for total dissolved solids contents. (8)ensure that detailed mineralogical analysis is carried out on rock samples from locations where key groundwater samples have been extracted.
*; *; *
JAERI-Data/Code 99-016, 183 Pages, 1999/03
JAERI-Data-Code-99-016.pdf:4.87MB
no abstracts in English
Kubo, Hiroatsu*; Omine, T.*; Yoshida, Kiyoshi
JAERI-Tech 98-050, 69 Pages, 1998/11
no abstracts in English
Honda, Yoichiro*; Hasegawa, Kazuo; Ouchi, Nobuo; Kusano, Joichi; Mizumoto, Motoharu
Proc. of 1st Asian Particle Accelerator Conf. (APAC98), p.74 - 76, 1998/11
no abstracts in English
Fujisaku, Kazuhiko*; Ishibashi, Yuzo; Takeda, Seiichiro; ; *
PNC TN8410 98-115, 50 Pages, 1998/09
None
