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Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Maruyama, Shuhei; Yokoyama, Kenji; Sugino, Kazuteru; Nagaya, Yasunobu; Oki, Shigeo
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
Ikeda, Kazumi*; Homma, Yuto*; Moriwaki, Hiroyuki*; Oki, Shigeo
Proceedings of 2014 International Congress on the Advances in Nuclear Power Plants (ICAPP 2014) (CD-ROM), p.1175 - 1183, 2014/04
Minato, Kazuo; Morita, Yasuji; Tsujimoto, Kazufumi; Koyama, Shinichi; Kurata, Masaki*; Inoue, Tadashi*; Ikeda, Kazumi*
Proceedings of 11th OECD/NEA Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation (Internet), p.341 - 349, 2012/00
In order to provide a quantitative assessment for the maturity of the partitioning and transmutation technology relative to its full-scale deployment, a technology readiness level (TRL) process was used. The definitions of TRL used in this study were based on those used in the Global Nuclear Energy Partnership (GNEP). The TRL was evaluated and the technology pathway was discussed for the systems of FBR and ADS for the minor actinides (MA) transmutation, MA partitioning processes, and MA-bearing fuels. Through the evaluation, it was recognized that hard requirements to be satisfied were present at TRL 5 for each technology development. The introduction of lab-scale tests with actual spent fuel for MA partitioning process and with actual separated materials for MA-bearing fuels fabrication and irradiation before the engineering scale tests may be effective and efficient solution.
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.
Hino, Ryutaro; Kaminaga, Masanori; Ishikura, Shuichi*; *; *; *; *; *
Proc. of 14th Meeting of the Int. Collaboration on Advanced Neutron Sources (ICANS-14), 1, p.278 - 287, 1998/00
no abstracts in English
*; *; *
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.
*; *; *
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.
Morii, Tadashi*; *
PNC TJ9214 90-002, 93 Pages, 1990/04
In the experimental fast reactor "JOYO", PNC (Power Reactor and Nuclear Fuel Development Corporation) schedules to move one control rod from the inner third row of the core to the outer fifth row. Two topics have been studied in order to get a license for the shift of control rod. Firstly, the work energy generated from expansion of the disruptive core material after the hypothetical core disruptive accident have been calculated by the VENUS code. The results show that the work energy of the core after the shift of one control rod increase by about 5MJ to 78 MJ compared with that of the core before the shift, but is still smaller than 120 MJ of the work energy described in the present documentation for petition of a license. Secondary, the effect of the reactor scram under the condition of the two rods stuck has been analyzed to examine a decrease of the safety margin of the scram worth. The calculated results of the HARHO-IN code shows that the consequences of the representative 4 accidents which are described in the present documentation for petition of a license are acceptably small.
Homma, Yuto*; Moriwaki, Hiroyuki*; Oki, Shigeo; Ikeda, Kazumi*
no journal, ,
no abstracts in English
Homma, Yuto*; Ikeda, Kazumi*; Oki, Shigeo
no journal, ,
no abstracts in English
Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Oki, Shigeo
no journal, ,
no abstracts in English
Ikeda, Kazumi*; Kan, Taro*; Maruyama, Shuhei; Ohgama, Kazuya
no journal, ,
no abstracts in English
