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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.
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.
*; *; *
PNC TJ9678 97-003, 80 Pages, 1997/02
In order to confirm the influence of lower U-235 enriched fuel on MK-III core, achievable U-235 enrichment is evaluated. The Pu enrichment, the fuel volume fraction, the structure volume fraction and etc. are chosen to be parameters. (1)Nuclear calculation of lower U-235 enriched core. Supposing enhancing the Pu enrichment, increasing the fuel volume fraction, reducing the structure volume fraction, extending the core height, employing N-15 enriched fuel and changing the Pu isotope ratio, the burnup calculation is performed so that the conditions of criticality and power distribution are satisfied and burnup characteristics and power characteristics are evaluated. Among the result, the linear heat rates are almost the same as those of MK-III standard core. The maximum of these burnup reactivity swing is increasing by 13%, the maximum of these fuel element burnup is increasing by 1% and the maximum of these fast neutron flux is increasing by 7%. (2)Calculation of U-235 enrichment. When the Pu enrichment of the outer core fuel is changed from 28.8w/o to 35w/o, the U-235 enrichment is reduced from 18.0w/o to 8.5w/o. Reducing structure volume fraction doesn't result in the reduction of the U-235 enrichment and increasing fuel volume fraction by 8% result in 13w/o of U-235 enrichment. When the core height extends from 50 cm to 60cm, the U-235 enrichment was reduced to 12%. Employing N-15 enriched nitride fuel lower the U-235 enrichment up to 5w/o. Supposing a Pu isotope ratio of weapon class, 9w/o of U-235 enrichment is feasible. Furthermore if the Pu isotope ratio is the weapon class and the Pu enrichment of outer core is increased to 33.4w/o, degraded U can be used.
*; *
PNC TJ9678 97-001, 234 Pages, 1996/03
A vertical seismic isolation system is composed of coned disk springs and lead dampers, and it is planed to apply to the common-deck-isolation-system in the large scale FBR Plant. The design study on the isolation system has been carried out, and the behavior of coned disk springs are tested. The results are as follows. (1)The lead damper, which absorbs the seismic energy by its plastic deformation, is applicable in the isolation system. (2)The isolation system is very effective when the natural frequency is less than 2.5 Hz. The displacement and the acceleration response are not sensitive to damping coefficient of the isolation device. (3)The rocking response of the common deck is very small. (4)The connection device for each coned disc spring using 'Connection-Ring' are designed. (5)The matelial called "WEL-TEN" series are applicable for the coned disk spring of the isolation device.
*; *; *
PNC TJ9678 96-010, 43 Pages, 1996/03
This work is an evaluation of reactivity curve of a boron-added testing model for Self-Actuating Shutdown System(SASS). The contents of this report are as follows. (1)Sample reactivity of boron and stainless steel. Two-dimensional RZ direct transport calculations of boron reactivity are done on condition that boron sample is loaded in the third row of the core. The difference or reactivity worth of boron among calculation methods is small and the reactivity worth of boron is negative in all axial positions. (2)Analysis of reactivity curve of testing model with boron for SASS. Several structures of testing model are given and their reactivity curves are calculated. In one testing model boron is added homogeneously in "meat section" of testing model and in the other testing models boron is added homogeneously in the down part of "meat section". Inserting the testing models from full-out position to full-in position, a negative reactivity of the former is bigger than one of the latter by a factor of l.52.0. In the other hand, inserting the testing models from halfway position to full-in position, no positive reactivity appears in the former but a small positive reactivity does in the latter. In conclusion, the operation testings with the boron-added model can be done without no positive reactivity, even if taking into account of uncertainty.
Pu contented fuel*; *; *
PNC TJ9678 96-009, 57 Pages, 1996/03
In this investigation, Pu fissile coefficients (reactivity ratio of nuclide) of MK-III core were calculated and Pu enrichment of three kinds of Pu composition were adjusted so that their reactivity worth are as much as ones of the fuel of MK-III standard core and the characteristics of MK-III cores with these fuels were evaluated. The contents of this calculation are as follows. (1)Calculation of Pu fissile coefficients. Normalizing coefficient of 
Pu as 1.0, Pu fissile coefficients (reactivity ratio of nuclide) of MK-III core were calculated about U, 
U, 
U, Pu, Pu, 
Pu, Pu and Am. The coefficients of U and Pu are 0.7 and 1.3. (2)Survey of fissile enrichment. Using Pu produced from spent LWR fuel of 60,70 and 80 GWd/t, as fuel of MK-III core, their enrichments of outer core fuel are about 32%, 34% and 36%. The higher Pu fraction of Pu is, the smaller burnup reactivity is. Maximum of reduction of burnup reactivity is 0.02% 
k/kk'. Using Pu produced from high burnup spent fuel, maximum linear heat rate is below 414 W/cm, maximum pin burnup is below 89,100 MWd/t. Power distribution and power peaking factor of these core are similar to ones of the MK-III standard core.