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Journal Articles

A Method for the prediction of the dose rate distribution in a primary containment vessel of the Fukushima Daiichi Nuclear Power Station

Okumura, Keisuke; Riyana, E. S.; Sato, Wakaei*; Maeda, Hirobumi*; Katakura, Junichi*; Kamada, So*; Joyce, M. J.*; Lennox, B.*

Progress in Nuclear Science and Technology (Internet), 6, p.108 - 112, 2019/01

In order to establish the prediction method of the dose rate distribution in the primary containment vessel (PCV) of the Fukushima Daiichi Nuclear Power Station, a series of calculations were carried out in the following way; (1) burnup calculation to obtain fuel composition at the time of accident, (2) activation calculation for the structural materials including impurities, (3) estimation of Cs contamination in PCV based on the result of severe accident analysis by IRID, (4) decay calculation of radioactive nuclides, (5) photon transport calculation to obtain dose rate distribution. After that, Cs concentration around the dry-well of 1F was modified to be consistent with locally measured dose rates in the PCV-investigation by IRID.

JAEA Reports

Evaluation of nuclear characteristics of minor actinide loaded core; Analyses of BFS-69 and BFS-66-2 critical experiments

Hazama, Taira; Sato, Wakaei*

JAEA-Research 2010-028, 101 Pages, 2010/09


This report describes analysis results on BFS-69 and BFS-66-2 critical experiments carried out under the collaboration with Russian Institute of Physics and Power Engineering. In the experiments, various nuclear characteristics were measured in 2 kinds of cores with/without Np loading of about 8 kg. JAEA's standard analysis results were presented with 4 kinds of nuclear data (JENDL-3.2, JENDL-3.3, JENDL/AC-2008, and ENDF/B-VII). The results show (1) An overestimation trend appears in BFS-69 criticality results, especially with JENDL-3.3 and JENDL/AC-2008. The difference from ENDF/B-II mainly lies in the average cosine of the scattering angle around 1 MeV; (2) A discrepancy exists in BFS-69 Na void reactivity results with the three JENDL nuclear data. The difference from ENDF/B-II mainly lies in scattering cross sections around 1 MeV and fission cross section around 1 keV; (3) The analysis results simulate measured Np effects on nuclear characteristics within experimental errors.

JAEA Reports

SLAROM-UF; Ultra fine group cell calculation code for fast reactor, version 20090113 (Translated document)

Hazama, Taira; Chiba, Go; Sato, Wakaei; Numata, Kazuyuki*

JAEA-Review 2009-003, 59 Pages, 2009/05


SLAROM-UF is a cell calculation code for fast reactors to produce effective cross sections with high accuracy in practical computing time, taking full advantage of fine and ultra-fine group calculation schemes. The fine group calculation scheme covers the whole energy range in a maximum of 900-group structure. Its group structure is finer above 52.5 keV with a minimum lethargy width of 0.008. Effective cross sections are evaluated based on the Bondarenko method. The ultra-fine group calculation scheme covers the energy range below about 52.5 keV. Its group structure is so fine ($$sim$$ 100,000 groups) as to treat resonance peaks as they are. Effective cross sections are calculated by solving an integral slowing down equation effectively, focusing only on elastic scattering and absorption reactions. Temperature can be specified freely by a user in the input data. The effective cross sections thus obtained are combined to calculate cell averaged cross sections.

JAEA Reports

SLAROM-UF; Ultra Fine Group Cell Calculation Code for Fast Reactor

Hazama, Taira; Chiba, Go; Sato, Wakaei*; Numata, Kazuyuki*

JNC-TN9520 2004-001, 97 Pages, 2004/03


A cell calculation code SLAROM-UF was developed to improve calculation accuracy of effective cross sections for various fast reactor types. SLAROM-UF has a capability to calculate effective cross sections in ultra fine groups of about 100,000 below 50keV and in fine groups above the energy (maximum 900 groups), Resonance interaction among the fuel, the coolant, and the structure materials can be treated accurately even in a heterogeneous cell structure. Temperature can be set up freely in a cell by the ultra fine group calculation. Improvement in nuclear characteristics was observed in the analysis of JUPITER critical experiment, as O.1% for criticality, 4% for sodium void reactivity, several % for radial reaction rate distribution, when SLAROM-UF was used insead of the typical cell calculation code. The effect of the ultra fine group calculation is remarkable in the non-leakage term of sodium void reactivity, and that of the fine group calculation is in the case that neutron spectrum in a core can not be represented by the cell calculation. When it is compared with a calculation by continuous energy Monte Carlo code MVP in a homogeneous lattice system, agreement lies within 1% for criticality, a few % for sodium void reactivity, and several % for radial reaction rate distribution of ZPPR-13A whose non-homogeneity is significant. The differences are reduced by about half, from those with the typical cell calculation code. SLRAOM-UF is easily available in the JOINT system currently being used in JNC, including all the functions available in the existent cell calculation code.

JAEA Reports

Results of Nuclear Design Accuracy Evaluation on BN-600 Hybrid Core

; Sato, Wakaei*; Hazama, Taira; Iwai, Takehiko*; Ishikawa, Makoto

JNC-TN9400 2003-074, 401 Pages, 2003/08


Nuclear design accuracy on the BN-600 hybrid core has been evaluated using the JNC's nuclear analysis system for FBR cores, by utilizing the critical experiment analysis results on BFS-62 configuration that had been obtained under JNC's efforts for Russian surplus weapons plutonium disposition. In the BN-600 hybrid core design, a part of the current UO2 fuel region is replaced by MOX fue1, and the Peripheral blanket region by stainless steel reflectors, respectively. These changes were simulated in a series of critical experiment configurations (BFS-62-1 to 4). Based on the analysis results on both BFS-62 configurations and other fast reactor cores, nuclear design accuracy on the BN-600 hybrid core has been evaluated by applying both the group constant adjustment method and the bias method. Evaluated nuclear parameters include, the criticality, fission rate distribution, sodium void reactivity, control rod worth, burn-up reactivity loss, etc. It is concluded, by applying the group constant adjustment method, that the evaluated accuracy (uncertainty) of most of the nuclear parameters can be decreased to less than half of those based on the basic nuclear constant without reflecting any experimental data. The improvement was mainly achieved by reducing the covariance of the iron elastjc cross section. This significant effect results from the feature of the BN-600 hybrid core, which has relatively larger power density, adopts U235 as the main fissile nucljde, and has the stainless steel reflector surrounding the fuel region. In addition, good consistency of analysis results between the BFS and other fast reactor cores is confirmed. Information obtained by BFS-62 experiment show significant contribution to the accuracy improvement. It is also found that the bias method shows less significant effects on the accuracy improvement than the group constant adjustment method. Furthermore, the bias method may degrade the accuracy for certain nuclear parameters that have large e

JAEA Reports

Integral test of JENDL-3.3 on Fast Reactors

Chiba, Go; Hazama, Taira; Sato, Wakaei*; Numata, Kazuyuki*

JNC-TN9400 2003-046, 49 Pages, 2003/05


An integral test has been carried out to evaluate a performance of evaluated nuclear data library JENDL-3.3, which was newly released, in a view of applying neutronics analyses of fast reactors. Japan Nuclear Cycle Development Institute has a large amoun

JAEA Reports

Evaluation of nuclear characteristics of minor actinide loaded core; An analysis of BFS-67 critical experiment

Hazama, Taira; Sato, Wakaei*; Ishikawa, Makoto; Shono, Akira

JNC-TN9400 2003-035, 44 Pages, 2003/05


Collaboration between Russian Institute of Physics and Power Engineering (IPPE) and Japan Nuclear Cycle Development Institute (JNC) named (Investigation of neutronic-physical characteristics and their change when introducing large quantity of neptunium (Np) at different BFS critical assemblies) is under progress. This is the first report of the collaboration to describe experimental information and JNC analysis results on BFS-67 critical experiment. In BFS-67 experiment, nuclear characteristics (criticality, control rod worth, sodium void reactivity, reaction ratio, etc) were measured in 4 different cores with various amounts of Np and location. JNC analysis was perfomed based on a JNC standard analysis scheme as in the analyses of BFS-62 critical experiments. (1)Sensitivity coefficients of Np capture cross section for the sodium void reactivity and control rod worth are large enough and comparable to those of U-238 and Pu-239. This indicates the experimental data can be used to improve design accuracy of Np loaded core. (2)C/E values for the criticality show high accuracy of 0.995 independent of core patterns, indicating accuracy of the calculation is high enough. (3)Calculated values for the sodium void reactivity agree with experimental values within 1cent and there is no influence of Np loading on calculation accuracy. (4)Calculated values for the control rod worth agree with experimental values within experimental errors for enriched B4C control rod. Those for naturaI B4C slightly overestimate. An influence of Np loading is not observed. (5)Calculated values for the reaction ratio agree with experimental values within 5% for fission reactions, whereas those for capture reactions show nearly 10% of differences. Positions of foils used in the measurement should be reflected.

JAEA Reports

Development of the unified cross-section set ADJ2000R for fast reactor analysis

Hazama, Taira; Chiba, Go; Numata, Kazuyuki*; Sato, Wakaei*

JNC-TN9400 2002-064, 315 Pages, 2002/11


ADJ2000R, the revised version of unified cross-section set ADJ2000, was developed. In ADJ2000R the error originated from JAERI FAST SET JFS-3-J3.2 is completely removed, which was not the case in ADJ2000. In the cross-section adjustment procedure, the error of JFS was completely removed from C/E (Calculation / Experiment) values and accordingly a different method was employed in evaluating analytical errors. Thereby degree of cross-section adjustment is largely different from that in ADJ2000, while change of C/E values by the adjustment are not affected. As a performance test, ADJ2000R was applied to a design analysis of 600MWe sodium cooled MOX fueled reactor. A drastic improvement was found in Doppler reactivity that was underestimated by about 10 % in ADJ2000 due to the error of JFS-3-J3.2. Differences from those with the original cross-section set are within a few percent except that burn up reactivity loss is 6% smaller. Design uncertainties are as small as those with ADJ2000 and are much reduced than those with the original cross-section set or the E/C bias method. ADJ2000R unified cross-section set has ability to predict accurately the various core characteristics of fast reactors from large cores to small cores, and from critical experiments to power reactors. ADJ2000R is open to the public, and is to be utilized in the feasibility study of future fast reactors.

JAEA Reports

Analyses on the BFS critical experiments; an analysis on the BFS-62-3A and 62-4 cores

Hazama, Taira; ; Iwai, Takehiko*; Sato, Wakaei*

JNC-TN9400 2002-036, 113 Pages, 2002/06


In order to support the Russian excess weapons plutonium disposition program, the intemational collaboration has started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engineering(IPPE). In the frame of the collaboration, analyses have been carried out for a series of critical experiments that simulate BN-600 (Russian commercial fast reactor). This report summarizes analysis results of the critical expeliments on BFS-62-3A and BFS-62-4 cores. BFS-62-3A core models BN-600 hybrid core in which the present BN-600 core is modified so as to partially load MOX fuel assemblies and replace the blanket region with stainless steel. BFS-62-4 core has the same layout as BFS-62-3A core except that the blanket region is not replaced. The analyses were performed with JNC standard method developed in the analysis of JUPITER experiment. The results show a good agreement with experimental values for the criticality and the reaction rate ratio. For the control rod worth and the reaction rate distribution, the results for BFS-62-4 core are also reasonable. However, for BFS-62-3A, analysis results overestimate the reaction rate in the stainless steel region by 20% and underestimate reactivity worth for one of the control rods by 10%. For the sodium void reactivity, underestimation of more than 20% were observed, but the disagreement were successfully solved by adopting a newly developed nuclear constant set with a fine group structure. In addition, analysis accuracies were compared among a series of analyses and it was confirmed that the introduction of MOX fuel assemblies does not affect the accuracy. The final goal of the work is to reflect the analysis results for designing BN-600 hybrid core. Then similarity was investigated between BFS-62-3A core and BN-600 hybrid core. A good similarity was found in the neutron spectrum, the fission reaction ratio, the fission reaction distribution, and the control rod worth. However, ...

JAEA Reports

Development of the unified cross-section set ADJ2000 for fast reactor analysis

; Numata, Kazuyuki*; Sato, Wakaei*;

JNC-TN9400 2001-071, 357 Pages, 2001/06


In the core design of fast breeder reactors, it is very impotant to improve the prediction accuracy of nuclear characteristics from the viewpoint of both reducing cost and insuring reliability of plant. The most powerfull method to reflect the C/E (Calculation/Experiment) values obtained from critical experimental analysis on the design work is the cross-section adjustment technique which is to unify cross-section covariance, integral experimental and analytical errors and the sensitivity coefficients of various cores and parameters based on the Bayesian parameter-estimation theory. The adjusted cross-section set is called "a unified cross-section set" here, since it combines integral experimental information with differential nuclear data. The main features of ADJ2000 compared with the preceding unified cross-section sets which were also developed by JNC (the former PNC) are as follows: First, the basic cross-section set adjusted was generated from JENDL-3.2, which is the latest version of the evaluated library in Japan at present. Second, the adjusted nuclear parameters include the self-shielding factors which were newly introduced in the adjusted parameters so that the accuracy of the Doppler reactivity can be improved. Third, the covariance data of nuclear parameters used in the adjustment were derived from the JENDL-3.2 Covariance File which has been completed and released by the Japan Nuclear Data Committee lately. Fourth, the integral experimental data were widely extended to include various independent facilities such as FCA in Japan, MASURCA in France, BFS-2 in Russia, JOYO as a power reactor, small core experiments in Los Alamos, as well as a series of JUPITER experiments in ANL/ZPPR that was only one experimental database in the previous adjustment study. Fifth, the integral data used in the adjustment includes the burnup- and temperature-related characteristics which are very important for power fast reactors. Finally, the statistical chi-square ...

JAEA Reports

Development of a standard data base for FBR core nuclear design (XIII); Analysis of small sample reactivity experiments at ZPPR-9

Sato, Wakaei*; Fukushima, Manabu*;

JNC-TN9400 2001-026, 90 Pages, 2000/09


A comprehensive study to evaluate and accumulate the abundant results of fast reactor physics is now in progress at O-arai Engineering Center to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as future commercial FBRs. The present report summarizes the analytical results of sample reactivity experiments at ZPPR-9 core, which has not been evaluated by the latest analytical method yet. The intention of the work is to extend and further generalize the standard data base for FBR core nuclear design. The analytical results of the sample reactivity experiments (samples: PU-30, U-6, DU-6, SS-1 and B-1) at ZPPR-9 core in JUPITER series, with the latest nuclear data libraly JENDL-3.2 and the analytical method which was established by the JUPITER analysis, can be concluded as follows: The region-averaged final C/E values generally agreed with unity within 5% differences at the inner core region. However, the C/E values of every sample showed the radial space-dependency increasing from center to core edge, especially the discrepancy of B-1 was the largest by 10%. Next, the influence of the present analytical results for the ZPPR-9 sample reactivity to the cross-section adjustment was evaluated. The reference case was a unified cross-section set ADJ98 based on the recent JUPITER analysis. As a conclusion, the present analytical results have sufficient physical consistency with other JUPITER data, and possess qualification as a part of the standard data base for FBR nuclear design.

JAEA Reports

Experimental analyses results on the BFS 58-1-I1 critical assemblies

; 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$$_{2}$$ 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$$_{2}$$ 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 ...

JAEA Reports

A Transplantation of the core parameter analytical system for fast reactors to the workstation environment (II)

Numata, Kazuyuki*; Sato, Wakaei*; ;

JNC-TN9440 2000-009, 127 Pages, 2000/05


Formerly a core parameter analysis was performed using the super computer environment in general. Recently a workstation has been equipped very widely, and in ordcr to utilize it for an analysis the transplantation of the core parameter analytical system is being carried out from the super computer into the workstation environment. This report describes the jobs performed for the transplantation and current situation of the compilation of the core parameter analytical system on the workstation environment. The document entitled "A Transplantation of the Core Parameter Analytical System for Fast Reactors to the Workstation Environment" reported the transplantation of the core parameter analytical system to the SUN workstation. This document represents the additional transplantation of the calculation code, which has not been transplanted yet to the SUN workstation and the transplantation from the SUN workstation to the DEC workstation aiming at the utilization of core parameter analytical system with the workstation other than SUN workstation.

JAEA Reports

ComparaUve analyses on nuclear charaderistics of water-cooled breeder cores

; Sato, Wakaei*;

JNC-TN9400 2000-037, 87 Pages, 2000/03


ln order to compare the nuclear characteristics of water-cooled bleeder cores with that of LMFBR, MOX fuel cell models are established for boiling and non-boiling LWR, non-boiling HWR and sodium-cooled reactor. Frst, the comarison is made between the heterogeneous cell calculation results by SRAC and those by SLAROM. The results show some differences as for neutron energy spectrum, one-grouped cross section and conversion ratio due to the different grouped cross section library (both are based on JENDL-3.2, though) used for each code, however, the difference is acceptably small for grasping the basic characteristics of the above-mentioned cores. Second, using the SLAROM code, main core parameters such as mean neutron energy, ratio of fast neutron and $$eta$$-value, are analyzed. The comparison between the cores show that softened neutron spectrum by the scattering effect of hydrogen or heavy hydrogen increase the contribution of nuclear reaction (especially for neutron capture reaction rather than fission reaction) in lower energy region comparing with LMFBR. ln order to overcome the effect, tighter lattice than LMFBR is necessary for water-cooled cores to realize the breeding of fissile nuclides. Third, effects of Pu isotopic composition on the breeding ratio are evaluated using SRAC burnup calculation. From the results, it is confirmed that degraded Pu (larger ratio of Pu-240) show the larger breeding ratio. At last, sensitivity analyses are made for k-effective and main reaction ratios. As for k-effective, using a temporary covariance data of JENDL-3.2, uncertainty resulting from the cross sections' error is analyzed for a boiling LWR and a sodium-cooled reactor. The boiling LWR core shows larger sensitivity in lower energy region than the sodium-cooled reactor (especially for the energy region lower than 1kev), And, 18-group analysis that is considered acceptably good for LMFBR analysis, should not be enough for accurate sensitivity estimation of ...

JAEA Reports

Development of a standard data base for FBR core nuclear design(X); Reevaluation of atomic number density of JOYO Mk-II core

Numata, Kazuyuki*; Sato, Wakaei*; ; Arii, Yoshio

JNC-TN9410 99-015, 136 Pages, 1999/07


The material composition of JOYO Mk-Il core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov.22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1)The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about $$pm$$0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis may be based on the isotope measurement of the manufacturing point of time without the decay of Pu-241. (2)As the other core components, the number densities of control rods and outer reflector-type A were largely improved.

JAEA Reports

Development of a standard data base for FBR core nuclear design, VIII; Compilation of JUPITER analytical results

Ishikawa, Makoto; Sato, Wakaei*; Sugino, Kazuteru; Yokoyama, Kenji; Numata, Kazuyuki*; Iwai, Takehiko*

PNC-TN9410 97-099, 512 Pages, 1997/11


A Standard data base for LMFBR core nuclear design has been developed to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as demonstration or commercial FBRs. To develop the data base, extensive work has been prerformed to accumulate and evaluate many kinds of results from fast reactor physics experiments and their analyses. The present report summarizes the analytical results of the JUPITER experiments, using the most recent nuclear data library (JENDEL-3.2) and the lates analytical methods in a consistent manner. In the present work, a great emphasis was placed on guaranteeing the essential requirements for this kind of general data base, that is, "Accountability", "Traceability" and "Consistency". In other words, consistent strategies and analytical methods were applied to all calculations, including detialed corrections; the enormous analytical input data generated were all saved in the form of computer files, so that reanalysis of any experiment could be easily performed for verification or in response to future improvement in nuclear data or analytical methods. The main results of the present JUPITER analysis are as follows: (1) The C/E(calculation/experiment) values of criticality were slightly underestimated by -0.7$$sim$$-0.3%$$Delta$$k. (2) The reaction rate ratio of C28/F49 was overestimated by +4$$sim$$+6% with the standard analytical method. However it was found to improve about 2% after the cell factors were revised using the Monte Carlo method. (3) The radial space-dependency of the reaction rate distribution and control rod worth almost disappeared in the homogeneous cores. (4) The previous overestimation of sodium void reactivity was greatly improved in the homogeneous cores.

JAEA Reports

Development of a standard data base for FBR core nuclear design(VII); Advances in JUPITER experiment analysis

Sugino, Kazuteru; Yokoyama, Kenji; Ishikawa, Makoto; Sato, Wakaei*; Numata, Kazuyuki*; Iwai, Takehiko*

PNC-TN9410 97-098, 247 Pages, 1997/11


The present report compiles the advances in experiment analyses of JUPITER, which was joint research programs between U.S.DOE and PNC of Japan, using the Zero Power Physics Reactor (ZPPR) large fast critical facility at ANL-Idaho in 1978 to l988. The advances here are use of the latest nuclear data library and the application of analytical methods which treat mechanisms in more detail or use fewer modeling approximations. As a result of using the latest nuclear data library, C/E values of nearly all characteristics approached unity, and the discrepancies between cores were reduced. Thus it is shown that the latest data library is effective for an analysis of nuclear characteristics. Further, an advance in analytical methods brought C/E value close to unity, and it clarifies the causes of differences between the calculational and experimental values. The current evaluation for each nuclear palameter shows following: (1)Criticality. The C/E values are from 0.993 to 0.997, a systematic underestimate. This underestimation is much smaller than the error caused by the uncertainty in nuclear data, which is the dominant error for this characteristic. In terms of analytical method, there are significant differences in calculation results between present and Monte-Carlo based methods, so more investigation will be required in future. (2)Doppler reactivity. The C/E values are from 0.8 to 0.9, a systematic underestimate. The analytical method, which is stood for by the use of ultra fine energy structure analysis, is so detailed that there is little room for improvement in that term. Therefore, some evaluation of the self-shielding factors and comparison with other Doppler reactivity experiments will be required. (3)Reaction rate distribution. It is judged that the present analytical method has an adequate accuracy for the core regions of homogeneous and axially heterogeneous cores, because the C/E values varied from unity by less than 2% for Pu-239 fission, U-235 fission ...

JAEA Reports

Development of a standard data base for FBR core nuclear design (V); Consistency evaluation of JUPITER experimental analysis

*; *; *; Sato, Wakaei*; *; Sanda, Toshio*

PNC-TN9410 95-214, 199 Pages, 1995/08


In order to improve the design method and accuracy of large fast breeder cores, extensive work has been performed to accumulate and evaluate many kinds of results of fast reactor physics experiments and analyses. As a part of efforts to develop a standard data base for LMFBR core nuclear design, the present report evaluates the physical consistency of JUPITER experimental analysis, especially concentrating on criticality. Here, the judgment of consistency is based on not only the deviation degree of C/E values from unity, but also various viewpoints such as the comparison with other cores or other nuclear characteristics by sensitivity analysis, the effect of changing nuclear data library, the analysis of FCA and JOYO which have completely different source of data from JUPITER, and the use of the Monte Carlo method as an analytical reference. (1)The C/E values of JUPITER criticality are slightly underestimated in the range of 0.993-0.999, using the JFS-3-J2 (1989) group constant set based on JENDL-2 and three-dimensional XYZ transport theory with the most detailed analytical model. There is an obvious dependency of C/Es on reactor core concepts with homogeneous or heterogeneous structure, the main cause of which is considered to be the effect of internal blanket existence and cross-section errors of JFS-3-J2, judged from sensitivity analysis. (2)The latest analytical method and model based on three-dimensional XYZ transport theory has sufficient ability to predict the relative changes of JUPITER criticality caused by the effect of reactor core size, CRP sodium channel, control rod and internal blankets. (3)The analytical error of JUPITER criticality was evaluated as approximately 0.3%dk and this seems reasonable, because the results of Monte Carlo analysis for ZPPR-9 criticality were almost identical with those of our standard analytical method. (4)The analytical results based on the latest JENDL-3.2 library were very close to those of JENDL-2 results, ...

Oral presentation

Influence of group constants on the Monju core physics test analysis

Takano, Kazuya; Sato, Wakaei; Usami, Shin; Nishi, Hiroshi

no journal, , 

no abstracts in English

Oral presentation

Monju core physics test analysis with various nuclear data libraries

Mori, Tetsuya; Sato, Wakaei; Uematsu, Mari; Hazama, Taira; Suzuki, Takayuki

no journal, , 

The previous Monju core physics test is analyzed using a JAEA's neutronics calculation system with various nuclear data libraries, JENDL-3.2, JENDL-3.3, JENDL/AC-2008, JEFF-3.1, and ENDF/B-VII. Analyzed the core characteristics are fixed absorber reactivity worth, fuel sub-assembly reactivity worth, coolant reactivity worth, burn-up coefficient, and reaction rate. It is found that the C/E (calculation over measurement) values are within experimental errors for the fixed absorber reactivity worth and the fuel sub-assembly reactivity worth. But that for the burn-up reactivity coefficient is around the experimental error and shows a tendency of overestimation. Further investigation is ongoing to improve the accuracy of the system. Obtained knowledge and experience will be reflected in the next physics test analyzes.

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