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Murakami, Masashi; Hoshino, Yuzuru; Nakatani, Takayoshi; Sugaya, Toshikatsu; Fukumura, Nobuo*; Sanda, Toshio*; Sakai, Akihiro
JAEA-Technology 2019-003, 50 Pages, 2019/06
JAEA-Technology-2019-003.pdf:4.42MB
Toward the establishment of a common approach to determine the radioactivity concentrations in dismantling wastes arising from research reactors, radionuclide concentrations in the reactor structure materials of aluminum, carbon steel, shield concrete, and graphite of TRIGA Mark II reactor at Rikkyo University, Japan, were evaluated with both radiochemical analysis and theoretical calculation. The measured nuclides by the radiochemical analysis were 
H, 
Co, and 
Ni in aluminum, H, Co, Ni, and 
Eu in carbon steel, H, Co, and Eu in shield concrete, and H, 
C, Co, Ni, and Eu in graphite. Neutron-flux distributions and neutron-induced activities were computed with DORT and ORIGEN-ARP codes, respectively. Using the results of material composition analysis, radioactivity concentrations were conservatively predicted with good accuracy except for graphite material.
Nagane, Satoru; Kitahara, Katsumi; Yoshikawa, Seiji; Miyasaka, Yasuhiko*; Fukumura, Nobuo*; Nishizawa, Ichio*
Dekomisshoningu Giho, (42), p.2 - 10, 2010/09
A reactor installed at the center part of the nuclear ship "Mutsu" has been stored safely and exhibited in a reactor room building since 1996. The reactor vessel and its internals are key components because of main radioactive wastes for the reasonable decommissioning plan in the future. This report describes the one-piece removal method as the one package of the reactor vessel with its internals intact with a shipping container or additional shields. The reactor vessel package (Max. 100 ton) will be classified acceptable for burial at the low level radioactive waste (LLW), which will be buried at a LLW pit facility under waste disposal regulations. And also, the package will be classified as an IP-2-equivalent package according to the requirement for Shipments and Packagings.
Takeuchi, Motoyoshi; Nakajima, Katsutoshi; Fukumura, Nobuo*; Nakayama, Fusao*; Ohori, Hideshi*
Dekomisshoningu Giho, (24), p.27 - 46, 2001/09
no abstracts in English
Iwashita, Mitsushige*; Arigane, Kenji; Kishimoto, Katsumi; Seiki, Y.*; Fukumura, Nobuo*; Mio, K.*
Journal of Nuclear Science and Technology, 37(Suppl.1), p.372 - 378, 2000/03
no abstracts in English
; Fukumura, Nobuo;
Nuclear Science and Engineering, 127(1), p.89 - 103, 1997/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)None
Kowata, Yasuki; Fukumura, Nobuo
PNC TN9410 96-131, 35 Pages, 1996/05
Plutonium fuel could be utilized in the entire core of heavy water moderated, boiling light water cooled pressure-tube-type reactor (HWR). The void reactivity, however, depends on the various parameters of the lattice. It is especially significant to clarify the influence of plutonium nuclides on the void reactivity. The void reactivities in the infinite HWR lattices have been parametrically analyzed to clarify the influences of changes in the lattice parameters on the void reactivity using the WIMS-D4 code with the JENDL-3.1 nuclear data. In this lattice calculation, it has been known that the behavior of the void reactivity can be made clear by separating the components for fuel nuclides, neutron cross sections, energy group and regions in lattice cell from the void reactivity using the important reaction rates. If the macroscopic 2200m/s neutron absorption cross section of fuel is identical each other, it has been shown that the void reactivity of the HWR lattice shifts further to the negative side in the narrower pitch lattice, and in the plutonium lattice than in the uranium lattice. The effect reducing the void reactivity to the negative by plutonium is caused mainly by the presence of the resonance cross section at around 0.3eV of 
Pu. Because the higher the content of Pu is, the less the recovery effect of neutron density within the resonance energy due to decrease in the thermal neutron scattering of hydrogen is with increase in coolant void fraction, so that the decreased resonance fission rate for Pu contributes to the more negative side for the void reactivity.
; Koyama, Tomozo; Funasaka, Hideyuki; Fukumura, Nobuo;
Nihon Genshiryoku Gakkai-Shi, 37(2), p.89 - 96, 1995/02
Times Cited Count：1 Percentile：17.53(Nuclear Science & Technology)None
; Fukumura, Nobuo
Nuclear Science and Engineering, 115(3), p.205 - 218, 1993/00
Times Cited Count：1 Percentile：18.76(Nuclear Science & Technology)None
*; *; Fukumura, Nobuo*; *; *; *; *
PNC TN1410 91-063, 239 Pages, 1991/08
no abstracts in English
Aihara, Nagafumi; Fukumura, Nobuo; Kadotani, Hiroyuki*; Hachiya, Yuki
Nuclear Science and Engineering, 109, p.158 - 170, 1991/00
Times Cited Count：4 Percentile：47.84(Nuclear Science & Technology)None
*; Fukumura, Nobuo*; *
PNC TN9410 90-179, 40 Pages, 1990/12
A new empilical fomula which serves for estimating the unknown subcriticality or test region in the two-region coupled reactor system has been formulated as an extention of the coupled reactor theory by R.Avery. This formula will be available for the two-region coupled reactor critical experiment with some reactivity perturbation to the driver region. The validity of this formula has been studied from the view points of reactivity perturbation means, suitable amount of reactivity perturbation, probable error due to the structure of the formula and so on through the computer simulation calculations. From the results it has been found that the treatment of the change of the coupling index is indispensable but the linear change model of the coupling index adopted in our formula is simple and effective. The rate of the change of the coupling index will be fairly different between the case of absorption perturbation and the case of water level perturbation, but it has been found that our formulas will be applicable to both cases within the allowable error by properly estimating the coefficient of the linear change of coupling index and by choosing the details of the reactivity perturbation to be suited to the cases. The maximum amount of reactivity perturbation suitable for the present method has been found to be 0.2 % 
K in the absorption perturbation method or 0.01 % K in the water level perturbation method. These values require fairly high precision measurement of boron concentration or of the level of moderator in the driver region, but we expect the experiments will be possible under careful execution of the measurement and suitable application of our formurlas.
Fukumura, Nobuo*
PNC TN9410 90-041, 116 Pages, 1990/03
Sub-criticality measuring experiment is planning for the purpose of contribution to safety and economy design of a nuclear fuel cycle facility This plan includes developing sub-criticality measuring methods and developing sub-criticality monitor Regarding as modification, the feasibility study was conducted in 1975. The adjustment design was conducted to decide the detail specification in 1988. Regarding as sub-criticality measuring minitor, fandamental study had been developed by the PNC/DOE Joint experiment from 1983 to 1988. This report discribes a work of DCA's modification, a study of the monitor, and the schedule for the experiment.
Kowata, Yasuki*; Fukumura, Nobuo
Nuclear Science and Engineering, 99(4), p.299 - 312, 1988/08
None
Fukumura, Nobuo*; *; *
PNC TN9410 88-072, 162 Pages, 1988/06
Improvement of calculational accuracy of ATR coolant void reactivity has been tried by using the more detailed calculational method as for the derivation of diffusion constant. The former calculational method of deriving the diffusion constant in WIMS-ATR code is the followings: (1)Calculation of collision probability with use of the model dividing the cell into three region (fuel, air gap and moderator). (2)Calculation of diffusion constant according to Bonalmi theory with use of the above calculated collision probability values. The present method is the one which calculates the collision probability more accurately by using multi region model including fuel, coolant, pressure tube and so on. For calculation of collision probability the CLUP code was used. By using the present code with the new calculational model of diffusion constant the coolant void reactivity were calculated. The following results were obtained: (1)The value of the diffusion constant is larger than the former one. This fact is more remarkable in the high voidage core and in the fast neutron energy region. (2)The present calculation accuracies of coolant void reactivity are 0.2%k for the DCA core and 0.07%k for the FUGEN core. These values are a half of the former ones. From the above results it is concluded that the present calculation model of the diffusion constant is effective for improvement of calculational accuracy of coolant void reactivity in ATR.
Fukumura, Nobuo*; *; *
PNC TN9410 88-030, 87 Pages, 1988/02
It is a difficult problem to shift coolant void reactivity of ATR to more negative. However it has been clarified by critical experiments with using DCA that use of the fuel pins containing gadolinia is very effective for decreasing coolant void reactivity. Then the analytical study has been done in order to find the solution which has the negative coolant void reactivity of ATR. The lattice is composed of the 36-rod fuel cluster and the lattice pitch is 24.2 cm square. The average enrichment through the MOX fuel cluster is 2.53 or 2.88 w/o Pufis.. This lattice is almost the same one as the demonstration power plant of ATR which is now under designing. The calculation code which is used is WIMS ATR - CITATION code system. The calculational input parameter is as follows; (1)Position of gadolinia poisoned fuel pin. (2)Number of gadolinia poisoned fuel pin. (3)Concentration of gadolinia contained in the fuel pin. (4)Concentration of 
B in the D
O moderator. (5)0, 20, 40, 60, 80 and 100% coolant void. The burnup calculation is up to 20 GWD/T. This is the core average value. The output of calculation is as follows: (1)Coolant void reactivity. (2)Reactivity coefficient of coolant void. (3)Local power peaking. (4)Decay curve of gadolinia concentration through burn up. According to the present study, the following understanding is clarified: (1)The best solution which has the negative coolant void reactivity value through burn up is the use of the 36-rod MOX fuel cluster which has the gadolinia poisoned fuel pin both in the inner and in the middle array of the cluster. (2)The number of the gadolinia poisoned fuel pin is 3 both in the inner and in the middle array. (3)Concentration of the gadolinia is 5 w/o. It is explained that the above effect is due to decreasigg the thermal neutron shielding of HO coolant by the strong neutron absorber gadolinia.
*; Fukumura, Nobuo*; *; Koyama, Tomozo; M.J.Hai*
PNC TN9410 87-072, 25 Pages, 1987/05
As part of a joint criticality data development between the power peactor and Nuclear Fuel Development Corporation (PNC) of Japan and the United State Department of Energy (USDOE), critical experiments have been conducted with organic moderated Fast Test Reactor (FTR) mixed oxide fuel pin arrays. The neutronic characteristics of an organic moderator can be examined by comparing the results of these experiments with the results of the same type of experiments performed with light water moderated system. In recent experiments performed at the Battelle Pacific Northwest Laboratories Critical Mass Laboratory, five distances of the lattice pitches were utilized which span from soft to hard neutron energy spectra. Results obtained by benchmark analyses of these experiments are discussed in this paper. The benchmark analysis was performed using the coupled WIMS-CITATION computer code whose accuracy has been demonstrated by applying them to heavy water reactor (ATR) with cluster type fuel. The scattering kernel used in the calculation for hydrogen atoms in light water was based on Nelkin's model. This same scattering kernel was used for hydrogen atoms in the organic moderator. The agreement between calculation and experiment (k-eff=1) for the light water moderated cores are fairly good (less than 0.2% k/k). On the other hand, k-eff's for the organic moderated core show a tendency to diverge from k-eff=1 gradually, as the lattice pitches become larger; the approximation used to calculate the scattering kernels for the organic moderator becomes less accurate. It is theorized that small difference in chemical bonding between light water and organic moderator account for the increasing deviation from k-eff=1 for the organic moderators, the larger the fuel pin lattice pitch. Future work would be necessary to establish a more accurate scattering kernel model of the organic moderator to precisely calculate critical conditions for organic moderated fuel pin ...
*; *; Fukumura, Nobuo*
PNC TN941 85-167, 66 Pages, 1985/11
Calculational accuracy of the WIMS-ATR code is analized based on post irradiation examination (PIE) data of Fugen MOX fuel and micro-parameter data of DCA fuel assemblies. In order to minimize discrepancies between calculational results by the WIMS-ATR and experimental ones, the nuclear library data in the WIMS-ATR are adjusted or added. The results are as follows. (1)The errors of heavy nuclide number density ratio by the WIMS-ATR burn-up calculation were max. 0.002% for uranium isotope ratio and max. 2% for isotope ratio at 16 GWd/t burn-up. (2)The error of 
Pu ratio was much decreased by adding Pu burn-up and product chain to the WIMS-ATR nuclear daTa library. (3)The effects of Pu or U cross section changes were calculated. The adjustments to the WIMS-ATR nuclear data library were proposed to decrease the error of Pu isotopic ratio to max. 1%. (4)The error of micro-parameter calculation was decreased by the adjustments.
