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Kamishima, Yoshio*; Yokoi, Shinobu*
JNC TJ9410 2005-002, 113 Pages, 2004/11
JNC-TJ9410-2005-002.pdf:5.84MB
This research considered formation nature examination of an apparatus vertical isolation system, and building arrangement of an apparatus vertical isolation system, while conducting investigation about the pliability of the element design accompanying cycle change of a vertical isolation element, and the possibility of design rationalization.
Kaise,Yoichiro*; Osada, Hiroo*
JNC TJ9400 2003-009, 183 Pages, 2003/03
JNC-TJ9400-2003-009.pdf:6.45MB
Various critical experiments have been analyzed and avaluated in Japan Nuclear Cycle Development Institute(JNC) to improve the accuracy of prediction for nuclear characteristics of fast breader reactors. This report describes update of the analysis of Monju Zebra Assembly Reactor Test (MOZART) reflecting a recent development of JNC analysis scheme. THe main results are as follows: (1)Compilation of Spectrum Measurements: Spectrum mesurement data are newly compiled including energy structure and geometrical information. (2)Reevaluation of atomic number density data: Atomic number density data were reevaluated considering impurities that had been neglected in the past analysis and reflecting a JNC standard analysis scheme. The revision of the data successfully reduces core type dependence of C/E values for criticality from 0.4%dk to 0.1%dk. (3)Analyses using JFS-3-J3.2R group constant set: The base-calculation and correction factors were fully reevaluated using JFS-3-J3.2R group constant set and the results were compared with those using JFS-3-J3.2. For criticality, C/E values become smaller by 0.1%dk, which tendency is consistent with that observed in the analysis of JUPITER experiment. Reduction of B-10 concentration dependence from 7% to 1% is observed in C/E values for control rod worth, and 10% improvement are for Na void reactivity. These improvements are attribute to the revision of the proup constant set and analysis scheme. The correction factors are confirmed to be insensitive to the revision of group constant sets.
*; *
JNC TJ9520 2001-002, 336 Pages, 2001/03
JNC-TJ9520-2001-002.pdf:7.28MB
The aim of this work is to develop insufficient auxiliary routines which manage input and output data and interface the main codes and to establish a shielding design analysis system on work stations (SUN, DEC). In shieldig design analyses, one- and two- dimensional (1-D and 2-D) transport Sn codes are used mainly with some auxiliary codes which generate input data of Sn calculation and edit Sn calculation outputs. The main transport calculation codes can be obtained from the Code Center of RIST (Research organization for Information Science & Technology). In this work. peripheral codes are developed to generate cross sections, produce Sn quadrature sets, edit calculation outputs or draw contour figures. In shielding calculations around a reactor, the boot-strapping technique is ofen employed to treat a large area extending from the core to the biological shield to improve the calculation accuracy. When a three-dimensional (3-D) calculation for a complex geometry with shielding defects, 2-D and 3-D coupling calculation is employed frequntly. To use this coupling method conversion codes are prepared which read flux file from DORT and prepare an external boundary source file for the 2-D or the 3-D calculation codes. For further conveniences well used data such as the Sn quadrature sets, the dose rate conversion factors, the reaction cross section sets are stored as a data base and code manuals including sample inputs of typical problems are prepared which are comprehensible to beginners.
*; Kishida, Masako*; *
JNC TJ9440 2000-013, 80 Pages, 2000/03
JNC-TJ9440-2000-013.pdf:4.93MB
In this work, analyses on sodium-water reaction temperaturc in the new SWAT-1(SWAT-1R) test were completed by the CHAMPAGNE code in order to understand void and velocity distribution in sodium system, which was difficult to be measured in experiments. The application method of the RELAP5/Mod2 code was investigated to LMFBR steam generator(SG)blow down analysis, too. The following results were obtained. (1)Analyses on sodium-water reaction temperature in the SWAT-1R test. (a)Analyses were carried out for the SWAT-1R test under the condition water leak rate 600 g/s by treating thc pressure loss coefficient, the interface friction coefficient and the coefficient related to reaction rate as parameters. The effect and mechanism of each parameter on the shape of rcaction zone were well understood by these analyses. (b)The void and velocity distribution in sodium system were estimated by use of the most suitable parameters. These analytical results are expected to be useful for planning of the SWAT-1R test and evaluation of test result. (2)Investgation of the RELAP5/Mod2 code. (a)The items to be improved in the RELAP5/Mod2 code were clarified to apply this code to the FBR SG blow down analysis. (b)One of these items was an addition of the shell-side (sodium-side) model. A sodium-side model was designed and added to the RELAP5/Mod2 code. Test calculations were carried out by this improved code and the basic function of this code was confirmed.
*; *
JNC TJ9440 2000-002, 90 Pages, 2000/03
JNC-TJ9440-2000-002.pdf:1.43MB
In order to support development of the dynamic reliability analysis program DYANA, analyses were made on the event sequences anticipated under emergency situations using the plant dynamics simulation computer code Super-COPD. In this work 9 sequences were analyzed and integrated into an input file for preparing the functions for DYANA using the analytical model and input data which developed for Super-COPD in the previous work. These sequences could not analyze in the previous work, which were categorized into the PLOHS (Protected Loss of Heat Sink) event.
Oyama, Kazuhiro*; Watanabe, Osamu*; *
JNC TJ9410 2001-002, 93 Pages, 2000/03
In the Strategic Research to Commercialize Fast Breeder Reactor Cycle plan, various breeder reactor core concepts are studied which are not restricted to the MOX-sodium combination. Metal and nitride are studied for fuels and gas, water, and lead for coolants. The objectives of this study is to compare the safety characteristics of the various breeder reactor cores by assuming the situation of the post-accident heat removal after hypothetical core disruptive accident. As a preliminary evaluation, coolable limit of core debris beds, which are formed after hypothetically disrupted core, was evaluated for the combinations of three types of fuels, MOX, metal and nitride, and four types of coolants, liquid sodium, lead, water and carbon dioxide gas. For the evaluation, a one-dimensional version of the DEBRIS-MD code which models the temperature distribution in a debris bed was used. Although the original code can handle only sodium coolant, special versions have been developed to handle lead, water and carbon dioxide gas coolants. Furthermore, the computer code for calculating debris bed temperature distribution was integrated in a newly developed coolant flow calculation model. It can handle arbitrary combination of coolant flow paths by using one dimensional flow network modeling. The computer code, named DEBNET was successfully used to analyze the post-accident heat removal in a 600MWe class FBR plant.
*; *
JNC TJ9440 99-022, 67 Pages, 1999/03
In the JNC design of the reference core of a commercial fast breeder reactor, priority is placed on high-burnup fuel and long-cycle operation leading to a low fuel cycle cost. Less attention is paid to the breeding characteristics, given the present situation of plutonium demand and supply. However, after about 2030, a rapid growth of fast breeder reactors is expected and the need for plutonium breeding is likely to arise. Therefore, we have studied feasibility of meeting a future demand for short doubling time, only by modifying the fuel assembly specifications of the reference core, when required after the commencement of the commercial plant operation. The survey has been made on a variety of modifications starting from the reference two-region homogeneous core under a fixed condition of average discharge fuel burnup of 150GWd/t. The study shows that an effective way to shorten the doubling time is a combination of modifications such as decreasing the core height and the gas plenum length, increasing the axial blanket thickness, increasing the number of fuel pins per assembly and shortening the period of the cycle operation. On the basis of the survey, the 39/-pin core has been chosen as a suitable candidate for short doubling time and the core characteristics have been evaluated. In addition, a study has been made on axial heterogeneous core configurations, revealing that the doubling times are about the same as that of the homogeneous core. This study shows that the complex inventory doubling time of the commercial plant can be shortened to about 30 years, only by modifying the fuel assembly specifications.
*; *; *
JNC TJ9440 99-021, 276 Pages, 1999/03
In order to support development of the dynamic reliability analysis program DYANA, analyses were made on the event sequences anticipated under emergency situations using the plant dynamics simulation computer code Super-COPD. The analytical models were developed for Super-COPD such as the guard vessel, the maintenance cooling system, the sodium overflow and makeup system, etc. in order to apply the code to the simulation of the emergency situations. The input data were prepared for the analyses. About 70 sequences were analyzed, which are categorized into the following events: (1)PLOHS (Protected Loss of Heat Sink), (2)LORL (Loss of Reactor Level )-J : failure of sodium makeup by the primary sodium overflow and makeup system, (3)LORL - G : failure of primary coolant pump trip, (4)LORL - I : failure of the argon cover gas isolation, and (5)heat removal only using the ventilation system of the primary cooling system rooms. The results were integrated into an input file for preparing the functions for the neural network simulation.
*; *; Suzuki, Katsuo*
JNC TJ9440 99-014, 73 Pages, 1999/03
Planning of the plutonium utilization in a thermal reactor have been investigated to evaluate the scenario for FBR development. Plans for the partial loading of MOX fuel in the Takaham-3,4 plant are studied. Information of the full MOX utilizing plans in an advanced light water reactor is summarized based on the documents distributed at the related technical committee. Nuclear compositions of the spent MOX fuel have been evaluated using SRAC and ORIGEN-2 code. Results of the study are as follows: (1)Surveying the status of MOX fuel utilization. Based on the public documents, the status and plans for utilizing both the partial loading and the full loading of MOX fuel in the LWR core have been summarized. (2)Evaluation of spent MOX fuel compositions. Nuclear compositions of the spent MOX fuel have been evaluated and summarized for both the APWR uranium core and APWR plutonium core.
U enriched fuels*; *; *
PNC TJ9678 98-003, 65 Pages, 1998/01
For the purpose of preparing a counterplan in the event that high U enriched uranium becomes difficult to secure, the characteristics of a lower U enriched MK-III core are evaluated. (1)Specifications of the Lower U Enriched Core. The specifications for three cases of the lower U enriched core are supposed. Under the condition that they are critical at the end of the equilibrium cycle and the power distributions are flater throughout the cycle, their U enrichment and Pu enrichment are determined as follows. Case 1:U enrichment 7.9w/o (outer core), Pu enrichment 35w/o. Case 2:U enrichment 5w/o (outer core), Pu enrichment 36.8w/o (outer core). Case 3:U enrichment 6.6w/o (outer core), Pu enrichment 29.8w/o. (2)Nuclear Calculation of Lower U Enriched Core. The results of nuclear calculation for lower U enriched core are shown as follows. (a)The criticalities of their cores are equal to that of an MK-III standard core. The maximum linear heat rates are increased from 414W/cm to 415W/cm. (b)The maximum fuel pin burnups are under 8.9
10
MWd/t. (c)The maximum fast flux increases to 4.210
/cm
s. (d)The flux spectrum shifts slightly toward the lower energy side. (d)In cases of weapon grade Pu, he isotope fractions of 
Pu and 
Pu double and the inventories of Pu fall by 14
15% at the end of fuel life.
*; *; Kishida, Masako*
PNC TJ9678 98-001, 294 Pages, 1997/09
The steam generator (SG) tube rupture phenomenon due to overheating by sodium-water reaction is considered as an important issue on SG safety evaluation and has been studied intensively. At this phenomenon, the cooling effect by the water/steam flow inside the tubes plays a significant role. Therefore, it is important to define the cooling effect by analyzing the behavior of the water/steam side during normal operation and during water/steam blow-down for overheating failure evaluation. In this work, the cooling effect was analyzed by a FBR SG blow-down analysis code, BLOOPH, and was corrected by taking the generated heat from the sodium-water reaction into account. In these analyses, the capacity and the operation method of the SG blow down system were treated as parameters. In order to confirm validity of the BLOOPH code, a similar analysis was carried out for the reference case by the thermal-hydraulic analysis code, RELAP5/Mod.2, that has been used widely for analyses of the LOCA phenomena of LWRs. The following results have been obtained by this work. (1)The effect of the capacity of the SG blow-down system on the SG blow-down characteristics has been well understood. A method has been found for reducing the time duration of the small flow rate which might occur inside the tubes during the blow-down. (2)A methodology has been established to design the most optimum SG blow-down system. (3)Analyses have been perfomed to define the cooling conditions needed for overheating failure evaluation. (4)The results by the code BLOOPH and RELAP5 have shown a reasonably good agreement regarding the water/steam pressure and the hydraulic behavior during the blow-down for the reference blow-down system. The validity of the BLOOPH code has been confirmed. (5)Research and development items have been clarified to improve the BLOOPH code in future.
*; *
PNC TJ9678 97-010, 184 Pages, 1997/03
A design study for a large scale seismic isolation FBR plant has been carried out to reduce the construction cost and to standardize the seismic design. This paper describes the studies of the vertical seismic isolation system for a common-deck which is applied in a horizontal isolated reactor building. The design have been studied on isolation device and absorber, and the characteristics of coned disk spring and its material strength are tested. The results of this study are as follows: (1)Modification and optimization of the coned disk spring, which is applied in the isolation device, has been carried out. (2)The integrity of support structures in the isolation device is clarified under the seismic loadings. (3)
-shaped lead dampers are applicable for the absorber system. (4)Nonlinearity of the lead dumper decreases the responsive displacement, but elastic stiffness of the dumpers increases the responsive acceleration, when too many lead dumpers are applied. (5)The rocking displacement and the amplitude of sloshing in the reactor vessel are very small, even if the imbalance of the weight of common-deck occurs. (6)The relationship between displacement and reaction force of the coned disk spring have been examined. In these tests, the spring stiffness significantly increases in the high-loading condition, because the edge of the coned disk spring bits into support plate.