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Oki, Shigeo; Oki, Shigeo
JNC TN9400 2003-041, 34 Pages, 2003/05
JNC-TN9400-2003-041.pdf:0.39MB
(ORIENT-cycle) is a fast reactor recycle concept based on removal of impedimental elements. With respect to the ORIENT-cycle examples developed in 2001, this report summarizes influences on core characteristics caused by the remaining fission products(FPs) in recycled fuel such as Zr,Mo,Pd,Cs,Ce,Nd,Sm,etc. In the case of one ORIENT-cycle example based on aqueous reprocessing, burnup reactivity (fuel cycle excess reactivity) of a conventional large fast reactor (sodium-cooled, MOX fueled) increases by about 0.7% delta k/kk' after the first recycle. For another kind of ORIENT-cycle example based on pyro-chemical reprocessing, burnup reactivity increases by about 0.3% delta k/kk' after the first recycle. Breeding ratio decreases by about 0.05 for the aqueous reprocessing case, and by about 0.02 for the pyro-chemical reprocessing case. Sodium void reactivity and Doppler constant also deteriorate, but these influences are small compared with burnup reactivity and breeding ratio. Due to the accumulation of FPs by multiple recycle, influences on the core characteristics become larger. Increment of burnup reactivity rises up to 2 % delta k/kk' for the aqueous reprocessing case, while 1 % delta k/kk' for the pyro-chemical reprocessing case. To reduce the above-mentioned significant influences, we need more efforts on increase of FP extraction coefficients and so on.
Hazama, Taira; Sato, Wakaei*; Ishikawa, Makoto; Shono, Akira
JNC TN9400 2003-035, 44 Pages, 2003/05
JNC-TN9400-2003-035.pdf:1.07MB
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.
Chiba, Go
JNC TN9400 2002-076, 32 Pages, 2002/12
JNC-TN9400-2002-076.pdf:1.45MB
Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate treatment of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of correction of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly.
Hazama, Taira; Chiba, Go; Numata, Kazuyuki*; Sato, Wakaei*
JNC TN9400 2002-064, 315 Pages, 2002/11
JNC-TN9400-2002-064.pdf:11.61MB
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.
Mikhail
JNC TN9400 2002-063, 53 Pages, 2002/11
JNC-TN9400-2002-063.pdf:0.33MB
This report is continuation of studying of the experiments performed on BFS-62-3A critical assembly in Russia. The objective of work is definition of the cross section uncertainties on reactor neutronics parameters as applied to the hybrid core of the BN-600 reactor of Beloyarskaya NPP. Two-dimensional benchmark model of BFS-62-3A was created specially for these purposes and experimental values were reduced to it. Benchmark characteristics for this assembly are (1)criticality; (2)central fission rate ratios (spectral indices);and (3)fission rate distributions in stainless steel reflector. The effects of nuclear data libraries have been studied by comparing the results calculated using available modern data libraries - ENDF/B-V, ENDF/B-VI, ENDF/B-VI-PT, JENDL-3.2 and ABBN-93. All results were computed by Monte Carlo method with the continuous energy cross sections. The checking of the cross sections of major isotopes on wide benchmark criticality collection was made. It was shown that ENDF/B-V data underestimate the criticality of fast reactor systems up to 2% 
k. As for the rest data, the difference between each other in criticality for BFS-62-3A is around 0.6% 
k. However, taking into account the results obtained for other fast reactor benchmarks (and steel-reflected also), it may conclude that the difference in criticality calculation results can achieve 1% k. This value is in a good agreement with cross section uncertainty evaluated for BN-600 hybrid core (
0.6% k). This work is related to the JNC-IPPE Collaboration on Experimental Investigation of Excess Weapons Grade Pu Disposition in BN-600 Reactor using BFS-2 Facility.
; Chiba, Go; Numata, Kazuyuki*
JNC TN9400 2002-062, 39 Pages, 2002/11
JNC-TN9400-2002-062.pdf:1.07MB
The unfied cross-section set ADJ2000, which was produced by adjusting 70-energy-grouped JFS-3-J3.2 (JENDL-3.2-based group constant set) using its analysis accuracy on various critical experiments, had been used in JFY2001 for core design studies in the feasibility study on commercialized fast reactor cycle systems. The ADJ2000 was published in June 2001, having 70 energy group structure as well. Calculated values from ADJ2000 would contain errors ("nuclear constant effects") which come from differences of weighting functions (neutron spectra) between that used for processing JFS-3-J3.2 and that of studied cores. In order to obtain more reliable results, the nuclear constant effects have to be corrected. Accordingly. nuclear constant effects on various fast reactor core concepts relating to application of unified cross-section set ADJ2000 have been evaluated for several nuclear parameters including criticality burnup reactivity loss, coolant void reactivity (depressurization reactivity in the case of gas-cooled reactor cores), Doppler reactivity and breeding ratio. Studied cores are listed below. (1)Sodium-cooled MOX-fuelled cores (large-size and middle-size) (2)Sodium-cooled metal-fuelled core (middle-size) (3)Lead-bismuth-cooled nitride-fuelled core (middle-size) (4)Carbon-dioxide-cooled MOX-fuelled core (large-size) (5)Helium-cooled coated-particle-fuelled core (large-size) (6)Helium-cooled enclosed-pin-fuelled core (large-size, 2 designs)
Chiba, Go
JNC TN9400 2002-057, 88 Pages, 2002/10
Effective cross sections obtained by cell calculations are used in core calculations in current deterministic methods. Therefore, it is important to calculate the effective cross sections accurately and several methods have been proposed. In this study, some of the methods are compared to each other using a continuous energy Monte Carlo method as a reference. The result shows that the table look-up method used in Japan Nuclear Cycle Development Institute (JNC) sometimes has a difference over 10% in effective microscopic cross sections and be inferior to the sub-group method. The problem was overcome by introducing a new nuclear constant system developed in JNC, in which the ultra fine energy group library is used. The system can also deal with resonance interaction effects between nuclides which are not able to be considered by other methods. In addition, a new method was proposed to calculate effective cross section accurately for power reactor fuel subassembly where the new nuclear constant system cannot be applied. This method uses the sub-group method and the ultra fine energy group collision probability method. The microscopic effective cross sections obtained by this method agree with the reference values within 5% difference.
Oki, Shigeo
JNC TN9400 2002-066, 110 Pages, 2002/07
JNC-TN9400-2002-066.pdf:14.26MB
Design study of core, fuel and related fuel cycle system with low-decontaminated fuel has been performed in the framework of the feasibility study (F/S) on commercialized fast reactor cycle systems. This report summarizes the influence on core characteristics of remaining fission products (FPs) in low-decontaminated fuel related to the reprocessing systems nominated in F/S phase I. For simple treatment of the remaining FPs in core neutronics calculation the representative nuclide method parameterized by the FP equivalent coefficient and the FP volume fraction was developed, which enabled an efficient evaluation procedure. As a result of the investigation on the sodium cooled fast reactor with MOX fuel designed in fiscal year 1999, it was found that the pyrochemical reprocessing with molten salt (the RIAR method) brought the largest influence. Nevertheless, it was still within the allowable range. Assuming an infinitetimes recycling, the alternations in core characteristics were evaluated as follows: increment of burnup reactivity by 0.5% k/kk', decrement of breeding ratio by 0.04, increment of sodium void reactivity by 0.1 
10
k/kk' and decrement of doppler constant (in absolute value) by 0.7 10
Tdk/dT.
Semenov, M.
JNC TN9400 2002-037, 58 Pages, 2002/06
JNC-TN9400-2002-037.pdf:2.51MB
This report is devoted to analysis of experimental studies performed on BFS-62-3A critical assembly in Russia. The objective of work is verification of the TRIGEX code for reactor neutronics analysis as applied to the hybrid core of the BN-600 reactor of Beloyarskaya NPP. Calculation models are described in the report, and results of analysis are compared with experimental data. The analysis was made by using the TRIGEX code mainly. FFCP code was used for taking into account heterogeneous structure of the BFS fuel regions. This code was coupled with the TRIGEX code for preparing averaged macro and micro cross-sections. Also, in the report, results of Monte-Carlo calculations with MMKKENO code are described. These calculations were used; first of all, for definition "reference" criticality and for confirming of correction values obtained with deterministic codes. The ABBN-93 system with the constant preparation CONSYST code was used as cross-section base. The following parameters were analyzed: criticality, control rod worth, sodium void reactivity effect, fission rate distribution and central reaction cross-section ratios (spectral indices). On the average, the differences between analytical results based on TRIGEX code calculations and experimental data do not exceed the following values: (1)0.1% k/k - for k
, (2)6% - for control rod worth, (3)4% - for fission rate distribution within the core, (4)0.2 pcm/kg - for sodium void reactivity effect This work is related to the JNC-IPPE Collaboration on Experimental Investigation of Excess Weapon Pu Disposition in BN-600 Reactor using BFS-2 Facility.
Hazama, Taira; ; Iwai, Takehiko*; Sato, Wakaei*
JNC TN9400 2002-036, 113 Pages, 2002/06
JNC-TN9400-2002-036.pdf:4.44MB
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, ...
Jin, Tomoyuki*; Oki, Shigeo
JNC TN9400 2002-015, 41 Pages, 2002/04
JNC-TN9400-2002-015.pdf:1.06MB
Transmutation of long-lived fission product (LLFP) using a fast reactor has been investigated in the framework of the feasibility study on commercialized fast reactor cycle systems in Japan. Neutron-moderating material is used in LLFP target subassembly for the sake of enhancing the transmutation efficiency, which causes the considerable changes of neutron distribution both in space and energy. At the present calculation the Monte Carlo method is employed for exact treatment of such a complicated system. Moreover the calculation method based on deterministic codes should be needed when we perform core design calculations and physical analysis. This study aimed at the establishment of a super cell modeling for LLFP target subassembly by using the general-purpose deterministic code system, SRAC95. In the fiscal year 2001, a super cell modeling has been worked out which appropriately deals with the incoming neutrons from the adjacent fuel region to the LLFP target, under the condition that the LLFP target is represented as homogeneous region.
Sugino, Kazuteru; ; Iwai, Takehiko*; Numata, Kazuyuki*
JNC TN9400 2002-008, 241 Pages, 2002/04
JNC-TN9400-2002-008.pdf:6.84MB
In order to support the Russian excess weapons plutonium disposition, the international collaboration has been started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engneering (IPPE). In the frame of the collaboration, JNC has carried out analyses on the BFS-62 assemblies that are constructed in the fast reactor critical experimental facility BFS-2 of IPPE. This report summarizes an experimental analysis on the BFS-62-1 and BFS-62-2 cores. The BFS-62-1 core models the present BN-600, and contains the enriched UO
fuel surrounded by the UO blanket. The BFS-62-2 core has the same layout as the BFS-62-1 but the blanket region was replaced with stainless steel shield. For core parameter analyses, the 3-D Hexagonal-Z or XYZ geometry model was applied by not only diffusion calculation but also transport calculation. Further in terms of the utilization of the BFS experimental analysis data for the standard data base for FBR core design, consistency evaluation with JUPITER experimental analysis data has been performed using the cross-section adjustment method. As the result of analyses, good agreement was obtained between calculations and experiments for the criticality, the reaction rate ratio and reaction rate distribution in BFS-62-1. In the reaction rate distribution of BFS-62-2 calculation without cross-section adjustment produced big radial dependency of calculation over experiment value (C/E value) in the core region and overestimation in the shield region. Cross-section adjustment technique procedure improved those estimation, however alternation of cross-section of Iron, which was dominant in above improvement, compared to the cross-section error, and further investigation was required. Concerning the control rod worth of BFS-62-1, radial dependency of the C/E value was observed whether cross-section adjustment technique was applied or not, therefore comparison with results of other BFS-62 ...
Chiba, Go; Numata, Kazuyuki*
JNC TN9400 2001-124, 44 Pages, 2002/02
JNC-TN9400-2001-124.pdf:1.29MB
It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL-3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.
Chiba, Go; Numata, Kazuyuki*
JNC TN9400 2001-109, 58 Pages, 2001/12
JNC-TN9400-2001-109.pdf:1.57MB
The fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL-3.2 has been widely used in the fast reactor analysis. However, it was recently found that there was a serious error in the process of applying the weighting function, a collision density spectrum in the inner core of "MONJU" as a representative of fast reactor spectrum. In this report, an effect of the error on nuclear characteristics was evaluated by a comparison with a new reactor group constant set which was produced by a standard reactor group constant set producing system of JNC with a correction of the weighting function. This report shows that the error of weighting function induces incorrect evaluation of neutron energy spectrum due to underestimation of scattering removal cross sections, and hence nuclear characteristics, such as criticality, sample Doppler reactivity, sodium void reactivity and reaction rate in a blanket region, are significantly affected. In addition, this report provides detailed information evaluated by separating "an effect of new generation reactor group constant" into more specific effects (i.e.an effect of correcting weighting function, an effect of differences in used cell codes and an effect of ultra fine group constant), which would be useful to investigate further on "an effect of new genaration reactor group constant".
Mantourov, G.
JNC TN9400 2001-106, 67 Pages, 2001/11
JNC-TN9400-2001-106.pdf:3.23MB
This work was fulfilled in the frame of JNC-IPPE Collaboration on Ex- perimental investigation of Excess Weapon Pu Disposition in BN-600 re-actor Using BFS-2 Facility. Data processing system CONSYST/ABBN cou- pled with ABBN-93 nuclear data libraty was used in analysis of BFS-62 and ZPPR JUPITER series fast reactor cores, applying JNC core calcu- lation code CITATION-FBR. FFCP cell code was used for taking into ac- count the spatial cell heterogeneity and response effect based on the first Flight collision Probability method and subgroup approach. Espe-cially two converting programs were written to transmit the prepared effective cross section to JNC standard PDS files to let then the CI- TATION code be applied for 3-D HEXZ neitronics calculation of the in- vestigated cores. The effects of nuclear data library have been stu- died by comparing the results calculated using ABBN-93 nuclear data library with the former ones obtained in JNC based on JENDL-3.2 nucle-ar data library. The
Oki, Shigeo; Jin, Tomoyuki*; *
JNC TN9400 2001-098, 154 Pages, 2001/10
JNC-TN9400-2001-098.pdf:4.33MB
Study for a conceptual image of LLFP (Long-Lived Fission Product) transmutation in a fast reactor has been started in the framework of the feasibility study on commercialized fast reactor cycle systems in Japan. Three species of LLFP such as iodine-129, technetium-99 and cesium-135 were preliminarily selected as the transmutation target nuclides, where the element separation process was assumed. The moderated-target subassemblies were used for loading LLFP on the reactor core. It is found that an effective transmutation is feasible for iodine-129 and technetium-99 by increasing the moderator fraction in the target subassembly, but extremely difficult for cesium-135 due to the emerging creation from the other isotope, cesium-133. Among the issues for future development, we have to pay careful attention to the linear heat rate for the iodine and cesium target pins whose melting point and thermal conductivity are considerably low.
Sugino, Kazuteru
JNC TN9400 2001-091, 75 Pages, 2001/08
JNC-TN9400-2001-091.pdf:2.21MB
In 1998, the feasibility study on FBR cycles has been started by Japan Nuclear Cycle Development Institute (omitted as JNC) and collaborative organizations in order to investigate practical Fast Breeder Reactor concepts of several types of coolants, fuel integrities and core arrangements. On the other hands, the conventional nuclear constant set is aiming at the analyses on sodium cooled MOX fuel cores only, that is, analyses on other types of cores are out of target. Therefore it can be said that an advance of the nuclear constant set system is essential for analyses on cores of various concepts. Under such a situation, JNC has compiled the constant set processing system that represented by NJOY and TIMS, and developed a new concept of nuclear constant set so called the next generation nuclear constant set for applications of various FBR analyses. The ordinary investigation has fixed the basic specification of the next generation nuclear constant set, and future study should decide its detailed specification. For the purpose of collecting information for the detailed design of the next generation nuclear constant set, new constant sets have been prepared using the constant set processing system, and applied to analyses on the JUPITER critical experiments, which are the mock-up critical experiments of sodium cooled MOX fuel cores. New information on the new constant set obtained is as follows: (1)The use of the new constant set improves the results of analyses on U-235 and Pu-239 fission reaction rate in the blanket region, sodium void reactivity and Doppler reactivity in comparison with those obtained using the conventional JFS-3-J3.2. (2)Application of the VITAMIN-J group structure reduces considerably errors of the calculated core parameters due to the choice of the weight function for the preparation of the nuclear constant set in comparison with that of the JFS-3 structure. Further it improves results of analyses on distributional core parameters a little bit.
; Numata, Kazuyuki*; Sato, Wakaei*; Sugino, Kazuteru
JNC TN9400 2001-071, 357 Pages, 2001/06
JNC-TN9400-2001-071.pdf:10.97MB
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 ...
Mantourov, G.
JNC TN9400 2001-069, 33 Pages, 2001/05
JNC-TN9400-2001-069.pdf:1.16MB
This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess of Weapon Pu Disposition in BN-600 Reactor using BFS-2 Facility. The data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS and ZPPR fast reactor cores applying JNC core calculation code CITATION. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the first flight collision probability method and subgroup approach. Especially a converting program was written to transmit the prepared effective cross sections to JNC standard PDS files. Then the CITATION code was applied for 3-D XYZ neutronics calculations of BFS and ZPPR JUPITER experiments series cores. The effects of nuclear data library have been studied by comparing the former results based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for ZPPR-9, ZPPR-13A and ZPPR-17A cores are presented. The calculated correction factor in all cases was less than 1.0%. So the uncertainty in C value caused by possible errors in calculation of these corrections is expected to be less than 0.3% in case of ZPPR-13A and ZPPR-17A cores, and rather less for ZPPR-9 core. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC culculation route revealed a large enough discrepancy in k-eff for ZPPR-9 (about 0.6%) and ZPPR-17A(about 0.5%) cores. For BFS-62-1 and BFS-62-2 cores such analysis is in progress. Stretch cell models for both BFS cores were formed and cell calculations using FFCP code have started. Some results of cell calculations are presented.
Dietze, K.
JNC TN9400 2001-043, 35 Pages, 2001/02
Sample reactivity measurements performed at the fast-thermal coupled facilities SEG / Germany and STEK / Netherlands have been re-analyzed using the JNC standard route for reactor calculation JENDL-3.2 // SLAROM / CITATION / PERKY. The SEG experiments comprise sample reactivity measurements with the most important stable fission products and structural materials in five reactor configurations with different neutron and adjoint spectra. The shapes of the adjoint spectra have been designed to get high sensitivities to neutron capture or the scattering effect. At the STEK configurations, the neutron spectra had an increasing neutron softness covering a broad energy range. A lot of FP nuclides have been measured. The calculated neutron and adjoint spectra are in good agreement with former results. C/E-values of the central reactivity worth (CRW) of about 90 materials (i,e., FP nuclides, structural materials, and standards) are given for 10 facilities. The C/E-values are compared with results obtained with the competitive European scheme JEF-2.2 // ECCO / ERANOS. For the SEG facilities, a crisscross use of JEF-2.2 with the JNC codes has been performed to get addional information about data and codes.