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Journal Articles

First ISTC/SAC seminar ``New approaches to the nuclear fuel cycles and related disposal schemes; Taking into account the existing excessive quantities of weapon-grade U and Pu, and reactor-grade Pu''

Maekawa, Hiroshi; Mukaiyama, Takehiko; Yamane, Tsuyoshi; *; *; Suzuki, Atsuyuki*; Takeda, Renzo*; *; Kawashima, Masatoshi*; *; et al.

Nihon Genshiryoku Gakkai-Shi, 40(12), p.963 - 965, 1998/12

no abstracts in English

Journal Articles

Development of a Method for Evaluation of Pin-wise Power Distribution in Fuel Assemblies of Fast Reactors

Yamaoka, Mitsuaki*; Kawashima, Masatoshi*; Yamaguchi, Takashi; Takashita, Hirofumi

Journal of Nuclear Science and Technology, 34(10), p.983 - 991, 1997/10

None

JAEA Reports

Module preparation for "MAGI" code system renewal (II)

*; Kawashima, Masatoshi*

PNC TJ9164 97-008, 245 Pages, 1997/03

PNC-TJ9164-97-008.pdf:6.96MB

"MAGI" system renewal has been continued for improving the prediction accuracy for neutronic and thermal characteristics along with the core-upgrading to the MK-III core. The present work made preparation of modules for neutronic calculations associated with adjoint flux, power and gamma-source distributions, burnup calculations. The input and output interfaces for the modules were also made for the full scale detail sample calculations under 24-mesh per subassembly diffusion model with 18 energy groups. The sample calculations provided good agreement in comparisons with the design calculation results. As for the future works, it needs to make several modules for heating cross section preparation, kinetic parameters, gamma flux, thermal characteristics, in addition to the I/O interfaces and system control modules, for the further renewals of "MAGI" system.

JAEA Reports

Preparation of unified cross section library for demonstration fast breeder reactor (III); Analysis of reactor physics experiment in "JOYO" startup test

Yamaoka, Mitsuaki*; Kawashima, Masatoshi*

PNC TJ9164 97-001, 185 Pages, 1997/03

PNC-TJ9164-97-001.pdf:4.94MB

To prepare unified library applicable to licensing calculation of the demonstration fast breeder reactor as well as base design calculation, the experimental data of the small reactor "JOYO" were re-evaluated and analyzed that had not been utilized in cross section adjustment before. The data of "JOYO" startup tests and operation were analyzed with cross section set based on the evaluated nuclear data library JENDL3.2 to provide integral data for cross section adjustment as power reactor data to supplement critical experiment data. The analysis is the first work that evaluated the core characteristics of "JOYO" MK-I and MK-II cores consistently. In the present work, errors and their correlation factors of both experiment and analysis data were evaluated for criticality, sodium void worth, fuel replacement worth and burnup coefficient. The recommended future work includes comparison of results with critical experiments, addition of experiment data to adjustment so as to reduce of dependency of prediction error upon core size.

JAEA Reports

Analysis of reduced sodium void reactivity core and improved Doppler reactivity core by utilizing the threshold reaction

*; Kawashima, Masatoshi*; *; Yamaoka, Mitsuaki*; Fujita, Reiko*

PNC TJ9164 96-008, 189 Pages, 1996/12

PNC-TJ9164-96-008.pdf:4.08MB

In the first reactor, sodium void reactivity and and sodium coolant temperature reavtivity are increased, and Doppler coefficient is decreased when MA is recycled. It is important to improve these reactivities in terms of safety features. In the study, Analysis was conducted of the effect on the sodium void reduction without deteriorating the core performance by mixing nuclides that give large absorption reactions in the higher energy region, where neutrons are increased at the neutron spectrum is hardened. Effect of higher Pu isotopes was also analyzed with parameters for improving Doppler coefficient. In the analysis of reduced sodium void reactivity core usig the threshold reaction, dominant energy region was identified for the sodium void reactivity by analyzing the core neutron spectra with parameters of MA mixing and core size. Furthermore, effect on reducing void reactivity was analyzed with parameters of inventory amounts and kinds of absorber nuclides. As a result, it was found that the sodium void reactivity of the MOX core with oxide-17 was about half of that of the core with natural oxide. In the analysis of reduced Doppler reactivity core, effect of improving the Doppler effect was analyzed for the nitride fuel core with higher Pu isotopes and resonance absorbers. Applicability and properties required for core analysis were also examined for candidates of base inert material, and the properties of fuel with those materials were preliminary surveyed. As a result, it was found that characteristics including Doppler coefficients can be improved with nuclides of structure material as metal form.

JAEA Reports

Study on advanced fast reactor; Analysis of burning Characteristics of long-lived FPs

Yamaoka, Mitsuaki*; Kawashima, Masatoshi*

PNC TJ9164 96-007, 106 Pages, 1996/12

PNC-TJ9164-96-007.pdf:1.96MB

Some of fission products (FPs) in spent fuels have very long half lives as transuranic nuclides. Fast reactors have a potential to transmute these FPs into short-lived ones because of high neutron flux. As neutron capture cross sections of FPs increase in low energy region, one of the effective means to transmute them is to irradiate them using target assemblies that contain pins loaded with neutron moderating material such as ZrH$$_{1.7}$$ as well as FP loading Pins. In the work carried out in 1994, a study was performed on transmutation of long-lived FPs, assuming that FP target assemblies are loaded in the core region of fast reactors. As a result, it was found that neutron moderation tends to improve transmutation rate whereas it causes to significant power spike in the adjacent core fuel. So as to suppress the power spike, FPs' loading at the outer periphery of the core was suggested. In the present study, an analysis was carried out to transmute FPs in fast reactors, assuming that target assemblies containing ZrH$$_{1.7}$$ pins and Tc pins are loaded at the outer periphery of the core. The analysis was mainly performed using the continuous energy Monte Carlo code that is effective for rigorous trestment of rcsonance absorption of Tc99 in the target assembly with large heterogeneity. The results are shown below. (1)To increase absorption rate of Tc99, the neutron spectrum where absorption in the resonance region is more than in the thermal region is advantageous. This is because the spectrum helps to suppress absorption by structural material. Therefore, the appropriate loading mass of moderator depends upon the moderating power. (2)The target assembly selected from the survey indudes 19 ZrH$$_{1.7}$$ pins with large diameter and 36 Tc pins with small diameter. The transmutation rate of Tc99 is about 4-5%/year and the mass is about 15-20kg/year. (3)The effect of loading target assembly upon main core characteristics were analyzed. It was found that the ...

JAEA Reports

None

*; Kawashima, Masatoshi*

PNC TJ9164 96-024, 137 Pages, 1996/03

PNC-TJ9164-96-024.pdf:3.33MB

None

JAEA Reports

A Preliminary analysis for transient safety tests in JOYO (II)

*; Kawashima, Masatoshi*

PNC TJ9164 96-017, 142 Pages, 1996/03

PNC-TJ9164-96-017.pdf:3.67MB

Accurate prediction methods for reactivity feedback coefficients such as power reactivity coefficients are most important to the core transiat test planning for safety capability demonstration using JOYO plant. This analysis studied components of feactivity feedback coefficients in the MK-III core, taking into accounts for fuel irradiation histories. The principal results are summarized as follows. Larger effects are expected to the reactivity feedback coefficients associated with fuel expansion and the Doppler components in the MK-III core than those in the MK-II cores, beause the MK-III core has the enlarged zones for high linear heat rating by by lateral zoning. This fact requires that fuel pin irradiation histories should be considered to the power coefficient prediction in the MK-III cores. The proposed model is appliable for the MK-III core reflecting experiences obtained in the MK-II core analyses, because the profiles and the component fractions of power reactivity coefficients in the MK-III cores are almost similar to those in the MK-II cores.

JAEA Reports

Grand design and module preparation for "MAGI" code system renewal

Kawashima, Masatoshi*; *

PNC TJ9164 96-012, 153 Pages, 1996/03

PNC-TJ9164-96-012.pdf:4.15MB

"MAGI" system renewal aims at improving the prediction accuracy for the neutronic and thermal characteristics along with the plant upgrade. Present work discussed the grand design of the system and prepared neutron flux distribution calculation module. The grand design study has determined that the system would cover pre-operation prediction function and post-operation recording function. It was proposed that the prediction calculations would enhance safety confirmation for power distributions and other major nuclear and thermal hydraulic chracteristics because the MK-III core would achieve power uprating with two-zoned regions. A prototypic flux calculation module was made with finite-difference 3-dimentional diffusion formulation for the Tri-Z geometry. The basic methodology for the new system includes 24mesh points per hexagon and 18 energy groups. The module was applied to the MK-III core as sample calculations. To proceed the "MAGI" system renewal, it is required to establish items to be compared with safety criteria using the neutronic module outputs. In addition, it needs to make the basic modules of the system for calculation control, constant preparation, various characteristics calculation, and IO control modules.

JAEA Reports

Nuclear and thermal analyses for the "MAGI" code system upgrade(2)

Kawashima, Masatoshi*; *

PNC TJ9164 96-003, 149 Pages, 1995/03

PNC-TJ9164-96-003.pdf:3.9MB

The core management code system "MAGI" was developed for the Experimental Fast Reactor, JOYO. The "MAGI" system upgrade aims at improving the prediction accuracy for the neutronic and thermal characteristics along with the plant up-grade. To achieve the goals, new appropriate methodology is needed for local characteristic predictions in the core and surrounging regions. The new basic methodology includes fine mesh-spacing (24 mesh poins per hexagon) and increased energy groups (18 neutron and 7 ga㎜a groups) with diffusion calculation. To utilize the fine mesh-spacing model with improved heterogeneous structures for varaious types of subassembly, simple rules are proposed for modeling, based on the numerical comparisons for reactivity worth, power distributions and burnup composition variations. To realize the "MAGI" system upgrade, it is required to establish further detail rules for various irradiation rigs for data flow specifications. It is also required to make the basic modules of the system, hereafter.

JAEA Reports

A Preliminary analysis for transient safety tests in JOYO (1)

*; Kawashima, Masatoshi*

PNC TJ9164 96-002, 207 Pages, 1995/03

PNC-TJ9164-96-002.pdf:4.48MB

As for demonstrating inherent safety features in fast reactors, transient safety tests will be planned using the experimental fast reactor JOYO. Accurate prediction for power reactivity-coefficients is most important to the core transients prediction. This work analysed the power coefficients measured in the initial and 27th cycles for the MK-2 cores. Major emphases were placed on the fuel temperature calculations and fuel axial expansion modelling proposed. The analyses explained that the observed systematic variations with core burnup were reproduced fairly well combining with two aspects mentioned above. These results have implied that further studies would incude adjustment methodology developement for the proper basic feedback model through various measurements as to establishing total reactivity feedback model to the Joyo core. Such methodology development efforts are required to the core bowing reactivity effects and feedback factors assosiated with the plant structure and oprational schemes, hereafter.

JAEA Reports

Core concept study on plutonium burning fast reactor (II)

Yamaoka, Mitsuaki*; *; Kawashima, Masatoshi*; Fujita, Reiko*

PNC TJ9164 95-009, 231 Pages, 1995/03

PNC-TJ9164-95-009.pdf:4.59MB

To enhance plutonium burning capability in fast reactors, one of the effective means is to use materials other than uranium for dilution of plutonium. A feasibility study was made to build a 600MWe-class core concept within the do-main of sodium-cooled fast reactors. The analysis covered core static and transient characteristics, including fuel material surveys. The candidate fuels were chosen as plutonium oxide with diluen materials, such as Al$$_{2}$$O$$_{2}$$ and BeO, to keep the Doppler coefficients negative large enough, condisering the TOP-type transisnts results from the FY1993 study. Core nuclear analysis showed that use of fuel without uranium considerably increases burnup swing and power mismatch between fresh and burnt fuels, aiming at the long cycle length as the 600MWe MOX core design. The core characteristics under ULOF- and UTOP-transients were compared with those in the 600MWe-MOX core. The study showed that the 9-month cycle core burned 59% fissile plutonium with negative sodium void worth (-1 $) under the plant condition for sodium inlet 390 C-deg. and the outlet temperature 510 C-deg. This study revealed that core neutronic feasibility has shown for such an innovative core concept with selecting appropriate diluent fuel materials combining core specifications. This means that sodium-cooled fast reactor has additional larger flexibility associated with plutonium utilization in the future.

JAEA Reports

A Parametric survey on advanced shielding subassemblies for reducing power generation in the IVST in the JOYO upgrade program

Kawashima, Masatoshi*; *; *

PNC TJ9164 89-002, 70 Pages, 1989/03

PNC-TJ9164-89-002.pdf:2.53MB

The JOYO-upgrade program has been planned for enhancing the present irradiation capacity with increase of the maximum power level and with improving its load factor. To achieve a higher load factor by reducing refueling intervals, the preseut work discussed a possibility of disusing the exclusive IVST pots in a invreased reactor power level. It is reqiured that power generation in the pot has to be suppressed to a level less than that can be coolable by natural convection, by replacing the present stainless steel reflectors to advanced B4C-shielding subassemblies. A parametric survey was made on some of a basic specification of the advanced shielding subassemblies as a concrete approach. The principal results obtained by this survey are as follows. (1)Use of the advanced shielding subassemblies can offer disuse of the exclusive IVST pots even with the upgraded power level up to 1.5 times larger than that of the present power level. (2)To achieve this target, two outermost layers of the reflectors are replaced by the advanced B4C-shielding subassemblies. This modification can be done within a slight alteration of the present flow distribution from lower pressure plenum as is done in the present version of the MK-2 core. (3)A 19-pin structure with 100cm height of the shielding materials in the axial length are proposed as a basie specification. Natural boron will be used for the 9-th row position and condensed boron-10 pellets will be installed for the 10th outermost row. These specifications are determined by effects of shielding ability, pellet maximun temperatures, PCMI and life of the subassembly. (4)The arrangement of the advanced shielding subassemblies makes the power generation of a fuel assembly in the IVST pots reduced considerably to the level which can be cooled by natural convection within a pot. But the power level is almost the same as the allowable upper limits of the power generation in a pot. Further discussion will be needed as for ...

JAEA Reports

JUPITER-III Experiment Analyses

Shirakata, Keisho; Nakashima, Fumiaki; Sanda, Toshio*; *; Kawashima, Masatoshi*; *; *

PNC TN2410 88-004, 359 Pages, 1988/03

PNC-TN2410-88-004.pdf:10.17MB

The JUPITER-III program, which started in Jan. 1987, successfully ended after one year experimental period without any delay from the original schedule. In the study last year, we mainly proceeded the experimental planning checks and data arrangement. This time studies were done putting an emphasis on the analysis of the experimental data and following conlusions were derived from the studies. (1)Pre-analysis for ZPPR-18 was done attempting to check the experimental plan. The result was actually reflected to the experiments for ZPPR-18 successfully. (2)Analysis of criticality for ZPPR-17A, 17B and 17C was done. The C/E value 1.0003 for ZPPR-17A comes to be almost the same value as that for ZPPR-9, which is 0.9995. (3)$$beta$$$$_{eff}$$ evaluation was made for ZPPR-17A and ZPPR-17B. The value obtained in this study is higher by 3 % than that of ANL, which shows the same tendency as found in the JUPITER-I and II analyses. (4)Control rod calculations were done for ZPPR-17A and 17B. The following C/E values were obtained as a result. Plate type control rod : 0.871$$sim$$0.899. Pin type control rod : 0.883$$sim$$0.890. (5)Experimental data arrangement was done for large zone sodium void experiments and for sodium void drawer oscillating experiments. (6)Criticality analysis for ZPPR-12 was performed for the pin cores and the plate Cores. The C/E value for central and edge pin zone cores agreed with those for the plate cores as shown bellow. Central pin : 1.018. Edge pin : 1.019. Plate cores : 1.020 (7)An attempt was made for building a data base system for the JUPITER programs, using the ZPPR-17A experimental data as an test example. (8)Multidrawer effect was evaluated for ZPPR-17A. The effect on criticality is +0.17$$Delta$$k/k.

JAEA Reports

Analysis of the start-up test of experimental fast reactor "JOYO"; The analysis of core nuclear characteristics (part-1)

Kawashima, Masatoshi*; *; *; *

PNC TN941 79-236, 238 Pages, 1979/12

PNC-TN941-79-236.pdf:9.39MB

This report describes the analysis of core nuclear characteristics of "Joyo" which were measured at the start up test. This report includes the following analyses; (1)Minimum Critical (2)Control Rod Worth (3)Sodium Void Reactivity (4)Fuel Subassembly Worth (5)Reaction Rate Distributions. The calculations were done by using JAERI-FAST-2 set (70 group). One of the purposes of this analysis was to confirm the validity of the 3-D Hex-Z model with diffusion theory. Followings are main results; (1)C/E value for K$$_{eff}$$ was 1.0044 in the minimum critical core. (2)C/E values of control rods worth were 0.97$$sim$$1.03. (3)The analysis of sodium void effect for single channel showed that the C/E values were 0.8$$sim$$1.0 in the core region and 1.3 at the blanket-core boundary with homogeneous model. The diffusion calculations with JFS-2 set showed good agreement between measured and calculated data. (4)The calculated substitution reactivity worth between core S/A and blanket S/A at the core periphery agreed well with the measured ones (C/E = 0.95$$sim$$1.05). The calculated substitution worth between Na channel and various S/A's generally overestimated the measured ones. (core fuel $$rightarrow $$ Na at the center C/E = 1.1$$sim$$ 1.2, blanket fuel $$rightarrow$$ Na at 5F1 C/E - 1.3$$sim$$1.5) (5)The analyses of the reaction rate traverses were done by the 3-D Hex-Z model with 8 group diffusion theory. C/E values of the central spectral indices for F$$^{49}$$/F$$^{25}$$, F$$^{28}$$/F$$^{25}$$ were 0.99 and 0.95 respectively. It was confirmed that 3D Hex-Z model was very useful to calculate whole core power distributions including the flux-tilt by the control rod insertion.

JAEA Reports

Start-up test of experimental fast reactor "JOYO"; Nuclear power calibration and power distribution (Part 1), NT-41 : Nuclear power calibraion, NT-42-1 : Power distribution (Part-1)

Yamamoto, Hisashi*; Sekiguchi, Yoshiyuki*; *; *; Kawashima, Masatoshi*; *; *

PNC TN941 79-112, 156 Pages, 1979/07

PNC-TN941-79-112.pdf:4.67MB

This report describes the results of the Nuclear Power Calibration test and the Power Distribution measurements made on the core center axis by micro fission chambers. The nuclear power leve1 was determined from the relationship between the calculated reaction rates and the count rates obtained by the Pu239 micro fission chambers which were inserted in the Joyo core. The Pu239 chambers used in this test were previously calibrated in the known neutron field of a thermal reactor. The principal results of this test are as follows: (1)It was confirmed that the relationship between the SRMS count rates and nuclcar power level is linear in the range of 0.1KW $$sim$$ 10 KW. (2)The SRMS and IRMS indicators overlap in the range of 1KW $$sim$$ 10 KW. The IRMS indicators show, linearity with the power level a above 1 KW. (3)Five fission reaction rates, Pu239, Pu240, U235, U238 and Th232, were measured along the center axis of the core using micro fission chambers. The axial power peaking factor was 1.19. The measured value of axial peaking factor agreed with the one predicted by design. (4)The SRMS count rates increase about 8% when temperature of primary sodium increases by 100$$^{circ}$$C. (5)The core fuel assemblies stored in the in-vessel fuel rack affect the count rates of the neutron monitoring system. When one core fuel assembly is loaded in a fuel rack position on a line connecting the core center and SRMS counters, the count rates increase about 25%. The additional nuclear characteristics determined in this teat are: (6)Isothermal temperature coefficient was -3.65 $$times$$ 10$$^{-3}$$ % $$Delta$$$$^{k}$$/$$_{k}$$/$$^{circ}$$C (190$$^{circ}$$C $$sim$$ 250$$^{circ}$$C) (7)The reactivity change due to the replacement of a normal core assembly at the core center with a special subassembly for this test was -0.085% $$Delta$$$$^{k}$$/k.

Journal Articles

None

Yamaguchi, Takashi; Kawashima, Masatoshi*

International Conference on Mathmatics and Computations,Reactor Physics,and Environmental Analysis, , 

None

Oral presentation

Reduction and resource recycling of high-level radioactive wastes through nuclear transmutation, 1-2; Investigation of system for volume-reduction and recycling of HLW

Nishihara, Kenji; Kawashima, Masatoshi*; Fujita, Reiko*

no journal, , 

In this project, research and development of elemental technologies has been promoted for separation and transmutation. In this research, we integrated HLW volume reduction and resource recycling system and evaluated the overall material balance. In addition, we examined the disposal method of generated waste and the possibility of recycled materials.

Oral presentation

Development of a passive safety shutdown device to prevent core damage accidents in fast reactors, 5; The Device structure concept and basic evaluations of fuel migration behavior in the device by visualization experiments

Sekio, Yoshihiro; Sato, Isamu*; Kawashima, Masatoshi*; Morita, Koji*

no journal, , 

A proposed passive reactor shutdown device contains pins with fuels that are kept in the solid state during normal operation but melt into the liquid when its temperature exceeds a prescribed value under severe accidents. The device leads reactor to subcritical state by liquid fuel migration to low positions in the pins. In this study, the device structure for fast reactor core was proposed based on safety evaluation results. The liquid fuel is needed to migrate before the accident occurs, and the migration time would depend on the fuel viscosity. To obtain basic acknowledge for determination of fuel chemical components, the effect of material viscosities on fuel migration time was evaluated through visualization experiments using liquid samples with different viscosities.

Oral presentation

R&D on mercury target for spallation neutron source to improve the durability under high power operation, 1; Mechanism of damage mitigation effects by gas bubbles and damage observation results

Kogawa, Hiroyuki; Kawashima, Hiroyuki; Ariyoshi, Gen; Wakui, Takashi; Saruta, Koichi; Naoe, Takashi; Haga, Katsuhiro; Futakawa, Masatoshi; Soyama, Hitoshi*; Kuji, Chieko*; et al.

no journal, , 

In a mercury target system of the J-PARC, an operation injecting microbubbles of helium gas into mercury is carried out to reduce the pressure waves that cause cavitation damage. It was confirmed the damage was mitigated by increasing the injection amount of gas bubbles, while the damage considered to be caused by impact pressure from the gas bubbles was observed. To improve durability, it is necessary to find the optimum bubble condition, and those are also important to evaluate the radiation damage of the vessel material and to develop a diagnosis technology. In this report, as the first report of the series, the outline of the development to improve the durability will be reported with the damage observation results.

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