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; *; *; *
JNC TN8410 2000-011, 185 Pages, 2000/05
JNC-TN8410-2000-011.pdf:4.67MB
This report describes the neutronic design calculational methods used in Fuel Design and Evaluation Group in order to inform other related sections of FBR core analysis technology and hand down the technology. Especially we show the neutronics calculation procedures used for the conceptual design study of the advanced core with 127 pin bundle for MONJU that has been carried out in our group. The topics include effective cross section preparation calculations, two-dimensional depletion calculations, three-dimensional diffusion calculations, reactivity coefficient calculations, and control rod worth calculations. The calculational methods shown in this report are the standard neutronics calculation methods employed in our group at the moment. However, the improvement of calculation codes, the reduction of correction factors and uncertainties for design using the nuclear data obtained in the start-up test for MONJU and so on, and the update of nuclear data file will be planned in order to improve evaluation accuracies. Those may change the neutronic design calculational methods, but we decided to describe the present standard calculational methods in our group from the viewpoint of sharing information in JNC.
; 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.
Takeda, Toshikazu*; Kitada, Takanori*; *; *
PNC TJ9605 98-001, 267 Pages, 1998/03
As the improvement of numerical analysis method for FBR core characteristics, studies on several topics have been conducted; multiband method, Monte Carlo perturbation and nodal transport method. This report is composed of the following three parts. Part 1: Improvement of Reaction Rate Calculation Method in the Blanket Region Based on the Multiband Method. A method was developed for precise evaluation of the reaction rate distribution in the blanket region using the multiband method. With the 3-band parameters obtained from the ordinary fitting method, major reaction rates such as U-238 capture, U-235 fission, Pu-239 fission and U-238 fission rate distributions were analyzed. As for the nuclides to be analyzed, the elements of structure material, such as iron, nickel, chrome and sodium were considered. By the present method, all the reactions became larger at the deep region in the blanket. The maximum correction amounted as much as 5%. This tendency lessen the disagreement between the ordinary calculation and the experiment. It was made clear that the treatment in inter-band scattering term is veryimportant because it has large sensitivity on the result. An alternative method to determine the multiband parameters whieh method is based on more direct approach and is free from drawbacks in the present method, was also investigated. Part 2 : Improvement of Estimation Method for Reactivity Based on Monte-Carlo Perturbation Theory. Perturbation theory based on Monte-Carlo perturbation theory have been investigated and introduced into the calculational code. The continuous energy Monte-Carlo perturbation code has been developed by using not only the correlated sampling method which is already used before, but also the derivative operator sampling method. The Monte-Carlo perturbation code was applied to MONJU core and the calculational results were compared to the reference. The change of eigenvalue caused by the change of sodium density in the GEM or dummy ...
Hazama, Taira
PNC TN9410 97-088, 139 Pages, 1997/10
Critical experiments were carried out on Deuterium Critical Assembly (DCA) modification core. DCA modification core has two regions, that is, test region and driver region. The test region consists of various types of fuel and moderator, while the driver region remains the same as the original DCA core (ATR simulated core). Critical characteristics were measured with various types of core patterns and were compared with calculated values based on SCALE code system. Monte calro code KENO was found to be very accurate in the core analysis. The accuracy stays below 0.5 %dk/k in keff even if core configulation is extremely complicated.
Takeda, Toshikazu*; *; Kitada, Takanori*; *
PNC TJ9605 97-001, 100 Pages, 1997/03
This report is composed of the following two parts and appendix. (I)Improvement of the Method for Evaluating Reactivity Based on Monte Carlo Perturbation Theory (II)Improvement of Nodal Transport Method for 3-D Hexagonal Geometry (Appendix) Effective Cross Section of 
U Samples for Analyzing Doppler Reactivity in Fast Reactors Part I. Improvement of the Method for Evaluating Reactivity Bascd on Monte Carlo Perturbation Theory. Theoretical formulation in Monte Carlo perturbation method had been checked, and then introduced into a calculation code. The increase of CPU time is about 10 to 20 % compared to that if normal Monte Carlo code, in the cases of same number of history. This Monte Carlo perturbation method found to be effective, because results are almost reasonable and deviations of the results are especially small, by using the Monte Carlo perturbation code. However, there are somc cases that the results of the change of eigenvalues becomes positive or negative by changing the estimator, and there is no reasonable difference in the results between the conventional method, which does not consider the change of neutron source distribution caused by a perturbation, and the new method, which consider that change. Thus it is still necessary to check the Monte Carlo pcrturbation code. Part II. Improvement of Nodal Transport Method for 3-D Hexagonal Geometry It is certain that we can accurately evaluate hexagonal geometry FBR core by nodal transport calculation code for hexagonal-Z geometry named 'NSHEX'. However it is also found that in very heterogeneous core the results is not good enough. Because the treatment of the transverse leakage to the radial direction, which is use for evaluating intra-nodal flux distribution, is not so accurate. For the treatment of the leakage distribution, it is necessary to estimate the nodal vertex flux. In conventional method, the vertex flux estimated by the surrounding node surfacc flux around that vertex. On the contrary,
Hayashi, Hideyuki; ;
PNC TN9410 96-062, 186 Pages, 1996/02
A conceptual design study on a 1300MWe large FBR plant was performed with focusing on enhancing passive safety and capital cost reduction. Spectrum-adjusted mixed nitride fueled core in which zirconium hydride was added, was applied to enlarge Doppler reactivity coefficient. Breeding ratio of 1.2 was obtained only with one layer of radial blanket subassemblies by optimizing the content of the zirconium hydride. The optimization also lightened the burden to the reactor structure through the reduction of the core diameter. Reactor passive shutdown were performed in the ATWS events of ULOF and ULOHS, and UTOP caused by one control rod full runout was endurable under the criterion of the prevention of coolant boiling. The safety feature can be called as inherent safety, because the feature comes only from the reactivity characteristics of the core. The integrities of the reactor structures which characterize head-access loop type reactor were evaluated on the transient thermal stress at the loss of flow accident and on seismic strain. Vertical strain of core support plate at loss of flow condition was also evaluated on the passive shutdown at ULOF. The capital cost of the large FBR plant was estimated 1.3 to 1.4 times as high as that of the same scale LWR based on the weight of major components.
; ; Uto, Nariaki; Yamaguchi, Akira; Kamide, Hideki; Ohshima, Hiroyuki; Hayashi, Kenji
PNC TN9410 94-235, 135 Pages, 1994/08
In this report passive prevention and mitigation measures with regard to core disruptive accident in future large scale liquid metal cooled fast breeder reactors are discussed and assessed. First the criteria for the assessment of passive safety measures are proposed, and the commonly proposed passive prevention and mitigation measures are briefly reviewed. Then innovative prevention and mitigation measures are newly proposed to provide additional mechanisms to limit the core damage or to prevent a recriticality event during the core disruption process. After assessing these passive measures based on the proposed criteria, appropriate combinations of the measures are recommended. Further, required R&D programs to confirm their effectiveness are described including necessity of a new in-pile experimental program.
Nakagiri, Toshio; ; Ohno, Shuji; ; *; Koyama, Shinichi; Shimoyama, Kazuhito
PNC TN9510 94-001, 246 Pages, 1994/05
None
*
PNC TN9410 93-010, 502 Pages, 1992/12
The present report compiles the experimental data of JUPITER phase-I, which was a joint research program between U.S.DOE and PNC of Japan, using the ZPPR facility at ANL-Idaho in 1978 to 1979. The JUPITER-I experiment was a series of critical experiments for conventional two-zone homogeneous cores of 600 to 800 MWe-class LMFBR, including seven experimental cores The nuclear characteristics recorded here include criticality, control rod reactivity, reaction rate distribution, sodium void reactivity, sample reactivity, Doppler reactivity, gamma heating and neutron spectrum. (1)ZPPR-9 : two-region cylindrical clean core with volume of app. 4,600 liters, (2)ZPPR-10A : hexagonal engineering-mockup core with 19 cotrol-rod positions(CRPs), (3)ZPPR-10B : changes seven CRPs to control rods(CRs) from ZPPR-10A, (4)ZPPR-10C : volume of app. 6,200 liters with similar core arrangement to ZPPR-10A, (5)ZPPR-10D : 31 CRPs with the same volume as ZPPR-10C, (6)ZPPR-10D/1 : changes the central CRP to a CR from ZPPR-10D, and, (7)ZPPR-10D/2 : changes seven CRPs to CRs from ZPPR-10D. The present work is a part of efforts to develop a standard data base for LMFBR core nuclear design at PNC. The detail of experimental data is thoroughly recorded here so as to re-analyze these experiments in future. In addition, these experimental data are installed in the computer system at OEC for convenience of analytical code input.
Yamaoka, Mitsuaki;
PNC TN9410 92-371, 94 Pages, 1992/12
TRU nuclides (Np, Am, and Cm) contained in the high level waste have extremely long-term radioactivity. They would be managed much more easily if transmuted in a short period. The present study deals with TRU transmutation by Fast Breeder Reactors (FBRs). The results are summarized below. (1)Study on a 300MWe Super Long Life Core for TRU Transmutation An FBR core loaded with TRU has a large potentiality of extending operation cycle length. Making use of the potentiality, a super long life FBR core loaded with TRU was studied aiming at continuous operation without refueling during plant life and efficient reduction of TRU nuclides. Core parameters were optimized with the electric power of 300MWe and analyses of nuclear and thermal characteristics were carried out. As a result, the burnup reactivity change of the optimized core for 34 years is very small (2.5% 
k/kk'). The power swing is also small, which resulted in satisfaction of the thermal design criteria. The amount of TRU transmuted during lifetime is about 5300Kg, which is equal to that 6 LWRs of 1000MWe produce during their lifetime. The Doppler coefficient (absolute value) is rather small because of TRU loading. Further study is needed on core kinetics from the view point of core safety and control. (2)Study on influence of uncertainties of TRU cross sections There are large uncertainties in TRU cross sections because of lack of experimental data. The influences of the uncertainties upon nuclear characteristics were evaluated for the super long life core and a large FBR core loaded with TRU of 5%. Sensitivity analysis on cross sections was carried out and uncertainties of nuclear characteristics were roughly evaluated. Based on the results, the TRU cross sections with large influences were identified.
Yamaoka, Mitsuaki; Hayashi, Hideyuki
PNC TN9410 92-368, 75 Pages, 1992/12
A passive safety test phase IIB is planned at the FFTF (Fast Flux Test Facility) core to assess the reactivity feedback effect related to passive safety feature of FBRs, especially the effect due to core deformation. For pre-test analyses of the test, a bowing reactivity analysis has been carried out for FFTF core. The bowing reactivity is analyzed based on core displacement data evaluated postulating ULOF (Unprotected Loss of Flow) event at 30% rated flow. In the analysis, fuel reactivity worth distribution is expressed as function on the reference core without deformation and the bowing reactivity is calculated based on the first-order perturbation theory. This report summarizes the relationships between power to flow ratio and the bowing reactivity with clearance between subassembly load pads and that between the core and the core restraint system as parameters. Followings are main results. (1)As the power to flow ratio increases, a positive reactivity is added to the core by the core deformation until clearance between subassembly load pads doses. This is due to the inward displacement of active core caused by mechanical interactions of subassemblies. (2)After the closure of clearance between subassembly load pads, the active core begins to move outwards, and a negative reactivity is added to the core. (3)The deformation behavior of the outermost subassemblies of the core dominates the bowing reactivity since both the magnitude of deformation and the reactivity effect for unit displacement are large compared with those of others. For the analysis, a code for bowing reactivity calculation has been developed. The calculation method and the manual are also presented in this report.
*
PNC TN9410 92-278, 347 Pages, 1992/09
A series of critical experiments for conventional two-zone homogeneous cores of 6 to 8 MWe-class LMFBR, JUPITER phase-I, were analyzed and evaluated using the latest analytical method, which had been established from the preceding numerous studies on fast reactor physics. The present work is a part of efforts to develop a standard data base for LMFBR core nuclear design at PNC. The analytical method and results are summarized as follows: (1)Analytical method (a)Nuclear data : 70-group fast reactor constant set JFS-3-J2(1989 edition) based on the Japanese Evaluated Nudear Data Library, version 2 (JENDL-2). (b)Cell calculation : plate stretch model, cell heterogeneity treatment by Tone's method and transport cross-sections weighted with neutron current. (c)Base core calculation : 18-group, three-dimensiona1 XYZ diffusion theory and Benoist's anisotropic diffusion coefficients. (d)Correction calculation : three-dimensional transport effect, mesh size effect, cell asymmetric effect and all master model effect etc. (2)Analytical results (a)The C.E (calculation/experiment) values of criticality agree quite well among seven cores (ZPPR-9
10D/2) and do not depend on the core volume or the number of control rod positions (CRP). (b)The C/E values of control rod worths increase gradually from the core center to the core edge positions in each core (511%). Those of reaction rate distributions also indicate similar spatial variations (25%), which is considered to be consistent with the C/E tendency of control rod worths. (c)The reaction rate ratios of C28/F49 and F25/49 give quite stable C/E values of 1.06 and 1.03, respectively. (d)The C/E values of sodium void reactivities are overestimated by +25% at core center region. The C/E dependence on void region size, which was pointed out in the past analyses, is found in the ZPPR-9 core, but not in the ZPPR-10 series. (e)The C/E dependence of 4% on the radial positions were found in sample ...