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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.
*; *; 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 97-003, 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 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.
*; Kishida, Masako*; *; *
PNC TJ9678 96-005, 86 Pages, 1996/03
The LMFBR concept without secondary sodium system is considered to be one of the most promising concepts for plant cost reductions. In this type of LMFBR, steam generators(SGs) are directly equipped in the primary sodium system. Therefore, the evaluation of effects of sodium-water reaction products (SWRPs) to the reactor core in case of SG tube rupture becomes one of the major safety issues. In this work, SMAC-13E, the thermal-hydraulic analysis code for sodium-water reaction developed by Power Reactor and Nuclear Fuel Development Corporation, was improved so as to evaluate the transport of SWRPs in the primary sodium system containing reactor core. In this improvement, SWRPs transport model (REACT), that treats solution, precipitation, deposition and so on, was added and it was intended that REACT is independent of the original part of SWAC-13E as much as possible. The improved analysis code (new SWAC-13E) was applied to the evaluation of SWRPs transport phenomena of typical planned LMFBR plant without secondary sodium system. The results of this analysis were qualitatively reasonable, and it was confirmed that the new SWAC-13E was applicable to the evaluation of the transport behavior of SWRPs.
*; *; *
PNC TJ9678 96-007, 133 Pages, 1995/11
Nuclear analyses are performed for the transition core to MK-III core. The contents of this calculation are as follows. (1)Excess reactivity or the transition core is 5.4 % k/kk' at the beginning of 35 cycle, which is below the nuclear limit, 5.5 % k/kk'. (2)Maximum linear heat rate is 355 W/cm, maximum fuel temperature is 2,298
C and maximum cladding temperature is 647C. These temperatures are below the thermal limits. (3) Minimum control rod worth of one rod stuck is 7.4% k/kk' at 32 cycle and 7.3% k/kk' at 35 cycle. The core or 100C is subcritical at one rod stuck. (4) The reactivity coefficients at 32 cycle and 35 cycle are near ones or MK-II core and MK-III core.
*; *; *
PNC TJ9678 96-004, 46 Pages, 1995/09
This calculation is evaluation of reactivity curve of a testing model of Self-Actuating Shutdown System(SASS) which gives data for application of permit of irradiation test in MK-III core of JOYO. The contents of this calculation are as follows. (1)Reactivity curve of testing model of SASS. Two dimensional RZ direct transport calculations are done on condition that the testing model is loaded in the radial center of core. Reactivity worth of the testing model of SASS is negative in the axial center region of core and positive in the region near the boundary between the core and the axial reflector. (2)Correction factor of reactivity worth of SASS for loading position. Correction factor of reactivity worth of SASS is calculated by two dimensional RZ transport code(TWOTRAN -II) and perturbation code(SN-PERT) because the testing model is planed to be loaded in the third row of core. The present structure of testing model is found to give 3 cent when it fall down from the full-out position.
Tada, Hiroyuki*; Kumasaka, Hiroo*; Saito, Akira*; Osada, Masahiko*; Maekawa, Keisuke
no journal, ,
no abstracts in English
Tada, Hiroyuki*; Saito, Akira*; Kumasaka, Hiroo*; Osada, Masahiko*; Takemura, Takato*; Maekawa, Keisuke
no journal, ,
no abstracts in English
Tada, Hiroyuki*; Saito, Akira*; Kumasaka, Hiroo*; Osada, Masahiko*; Takemura, Takato*; Maekawa, Keisuke
no journal, ,
no abstracts in English
Maekawa, Keisuke; Tada, Hiroyuki*; Saito, Akira*; Kumasaka, Hiroo*; Osada, Masahiko*; Takemura, Takato*
no journal, ,
no abstracts in English