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Journal Articles

Re-evaluation of assay data of spent nuclear fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems

Suyama, Kenya; Murazaki, Minoru*; Okubo, Kiyoshi; Nakahara, Yoshinori*; Uchiyama, Gunzo

Annals of Nuclear Energy, 38(5), p.930 - 941, 2011/05

 Times Cited Count:13 Percentile:23.58(Nuclear Science & Technology)

The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries. In this paper, postirradiation examination (PIE) data from two fuel assemblies irradiated in the Ohi-1 and Ohi-2 PWRs in Japan are shown. The PIE data from Ohi-2 has already been published. However, these data were not suitable for the benchmarking of calculation codes and libraries because several important specifications and data were not included. This paper summarizes the details of PIE data from Ohi-1 and Ohi-2 and shows in detail the data and specifications required for analyses of isotopic composition. For precise burnup analyses, the burnup values of PIE samples were re-evaluated in this study. These PIE data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. This indicates that the quality of PIE data from Ohi-1 and Ohi-2 PWRs is high, and that these PIE data are suitable for the benchmarking of burnup calculation code systems.

JAEA Reports

Measurement of neutron dose under criticality accident conditions at TRACY using ebonites

Murazaki, Minoru; Nobuhara, Fumiyoshi*; Iwai, Shohei*; Tonoike, Kotaro; Uchiyama, Gunzo

JAEA-Technology 2009-045, 46 Pages, 2009/09


Neutron doses under criticality accident conditions at TRACY were measured using ebnites, which are hard rubber containing sulfur. To evaluate a neutron dose, beta rays emitted from $$^{32}$$P induced by $$^{32}$$S(n,p) reaction are measured with Geiger-M$"{u}$ller (GM) counter. A calibration factor (Gy/cpm), which is pre-determined using a $$^{252}$$Cf source, is applied to the count rates to obtain neutron doses. Factors to correct for the difference between responses of $$^{32}$$S(n,p) to the spectrum of $$^{252}$$Cf source and to spectra of TRACY were calculated and applied to the doses. Ebonites were exposed by TRACY with and without the water reflector. Neutron doses in TRACY without a reflector were evaluated with an uncertainty of less than about 40%. On the other hand, average of neutron doses in TRACY with the water reflector were accurate. By these measurements, it was found that ebonites can be used as a neutron dosimeter for criticality accidents.

JAEA Reports

Re-evaluation of dose measurements under criticality accident conditions at SILENE and TRACY using TLDs

Murazaki, Minoru; Tonoike, Kotaro; Uchiyama, Gunzo

JAEA-Technology 2009-022, 49 Pages, 2009/06


We have re-evaluated the dose measurements at SILENE with TLDs for neutrons and with TLDs for $$gamma$$ rays, and at TRACY with TLDs for neutrons. The measurements with TLDs for neutrons were re-evaluated by revising factors used for calculation of doses from measured data. The re-evaluated results of TLDs for neutrons at SILENE agreed with the reference value given by IRSN within 10%. The re-evaluated results of TLDs for neutrons at TRACY are consistent with dose and distance from the surface of the core tank. The re-evaluated results at TRACY were about 50% larger than results of polymer-alanine dosimeters and those of TLDs for neutrons measured by the authors. The measurements at SILENE with TLDs for $$gamma$$ rays were re-evaluated by revising the method for obtaining doses from measurement data. By the re-evaluation, it was confirmed that the methods described in the present report are valid for processing measured data of TLDs for neutrons and those for $$gamma$$ rays.

JAEA Reports

SWAT3.1; The Integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP

Suyama, Kenya; Mochizuki, Hiroki*; Takada, Tomoyuki*; Ryufuku, Susumu*; Okuno, Hiroshi; Murazaki, Minoru; Okubo, Kiyoshi

JAEA-Data/Code 2009-002, 124 Pages, 2009/05


Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC widely used in Japan and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinide and the fission products in the spent nuclear fuel. Because of the ability to treat the arbitrary fuel geometry and no requirement of generating the effective cross section data, there is a great advantage to introduce continuous energy Monte Carlo Code into the burnup calculation code. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP and ORIGEN2. This report describes the outline, input data instruction and several example of the calculation.

Journal Articles

Measurement of neutron dose under criticality accident conditions at TRACY using TLDs

Murazaki, Minoru; Tonoike, Kotaro; Uchiyama, Gunzo

Journal of Nuclear Science and Technology, 46(2), p.193 - 203, 2009/02

 Times Cited Count:1 Percentile:88.61(Nuclear Science & Technology)

To develop a method for measuring neutron dose easily during criticality accidents, neutron ambient dose equivalents at the TRACY facility have been measured using neutron dose equivalent monitors with thermoluminescence dosimeters (TLD monitor). The TLD monitor is composed of two TLD badges and a cubical polyethylene case, and has a response similar to the ambient dose equivalent. In our experiments, TRACY was operated with and without water reflector to irradiate the TLD monitors. Measured ambient dose equivalents were proportional to the integrated power of TRACY, and agree well with calculation results of MCNP5. The measurement data were converted into tissue kerma using dose conversion factors calculated by MCNP5. Response correction factors to be applied to the measurement data considering the difference between responses of the TLD monitor to the $$^{252}$$Cf calibration source and to TRACY were also calculated by MCNP5. The neutron kerma ranged from 30 mGy to 15 Gy, which covers the range from 100 mGy to 10Gy specified as important in criticality accident dosimetry by the IAEA. The TLD monitor also satisfies the time limit on determination of doses required by the IAEA.

Journal Articles

Active reduction of the end effect by local installation of neutron absorbers

Suyama, Kenya; Murazaki, Minoru; Okubo, Kiyoshi; Okuno, Hiroshi

Annals of Nuclear Energy, 35(9), p.1628 - 1635, 2008/09

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

In the analysis of the burnup credit, it has been pointed out that the neutron multiplication factor becomes greater if we consider an axial burnup distribution of spent fuel assemblies than the case under an assumption of an average burnup through the fuel assemblies. This phenomenon is called "end effect" and it is one of the main technical issues in the burnup credit study. In this study, the reason why the end effect occurs in the criticality calculation of spent fuel assemblies is discussed by analyses of neutron flux distribution measurement both fixed source and eigenvalue calculations. These calculations show us that the end effect is induced by the solution of neutron balance equation as eigenvalue problem and an actual neutron flux increase occurs only when the neutron multiplication factor is close to unity. Based on the discussion, reducing the end effect actively by local installation of neutron absorbers (LINA) around the end regions of the fuel assemblies are proposed and its effect was confirmed based on the several criticality calculations.

Journal Articles

Decreasing in neutron multiplication factor in spent fuel storage facilities by changing fuel assembly position in axial direction

Suyama, Kenya; Murazaki, Minoru; Okuda, Yasuhisa*

Annals of Nuclear Energy, 34(5), p.417 - 423, 2007/05

 Times Cited Count:1 Percentile:87.77(Nuclear Science & Technology)

Long-term spent fuel (SF) storage becomes one of the options in nuclear fuel cycle strategies in Japan. In such facilities, criticality safety should be confirmed. Generally, to reduce the neutron multiplication factor, taking wide spaces between spent fuel assemblies is usually considered. However, from the view point that reducing neutron interaction between regions where contain much fissile materials, we are able to shift the assemblies in axial direction alternatively. In this study, the possibility of reduction of neutron multiplication factor by the alternative axial shift of spent nuclear fuel is presented. Several criticality calculations were conducted in order to evaluate the effect of alternative axial assembly shift (A3S) method, which shift the assembly not to horizontal but to vertical directions. For both PWR and BWR fuels, we have smaller neutron multiplication factors if we adopt A3S method than normal configuration. Concerning the end effect which is important for criticality of spent nuclear fuels, it was confirmed that A3S method reduces the end effect. This study shows that several configurations can be considered to optimize criticality safety design of spent fuel storage facilities to reduce the neutron multiplication factor and the end effect for effective operation of such facilities.

Journal Articles

Development of a criticality evaluation method considering the particulate behavior of nuclear fuel

Sakai, Mikio; Yamamoto, Toshihiro; Murazaki, Minoru; Miyoshi, Yoshinori

Nuclear Technology, 149(2), p.141 - 149, 2005/02

 Times Cited Count:1 Percentile:88.54(Nuclear Science & Technology)

In the conventional criticality evaluation of the nuclear powder system, the effects of particulate behavior have not been considered. In other words, it is difficult to reflect the particle behavior into the conventional criticality evaluation. We have developed a novel criticality evaluation code to resolve this issue. The criticality evaluation code, coupling a Discrete Element Method simulation code with a continuous-energy Monte Carlo transport code, makes it possible to study the effect of the particulate behavior on a criticality evaluation. The criticality evaluation code has been applied to the powder system of the MOX fuel powder agitation process. The criticality evaluations have been performed under mixing the MOX fuel powder in a stirred vessel to investigate the effects of the powder boundary deformation and particulate mixture conditions on the criticality evaluation. The evaluation results revealed that the powder uniformity mixture condition and the boundary deformation could make the neutron effective multiplication factor decrease.

JAEA Reports

Measurement and analysis of neutron flux distribution of STACY heterogeneous core by position sensitive proportional counter (Contract research)

Murazaki, Minoru; Uno, Yuichi; Miyoshi, Yoshinori

JAERI-Tech 2003-029, 107 Pages, 2003/03


We have measured neutron flux distribution around the core tank of STACY heterogeneous core by position sensitive proportional counter (PSPC) to develop the method to measure reactivity for subcritical systems. The neutron flux distribution data in the position accuracy of $$pm$$13mm have been obtained in the range of uranium concentration of 50g/L to 210g/L both in critical and in subcritical state. The prompt neutron decay constant, $$alpha$$, was evaluated from the measurement data of pulsed neutron source experiments. We also calculated distribution of neutron flux and $$^{3}$$He reaction rates at the location of PSPC by using continuous energy Monte Carlo code MCNP. The measurement data was compared with the calculation results. As results of comparison, calculated values agreed generally with measurement data of PSPC with Cd cover in the region above half of solution height, but the difference between calculated value and measurement data was large in the region below half of solution height. On the other hand, calculated value agreed well with measurement data of PSPC without Cd cover.

JAEA Reports

Derivation of correction factor to be applied for calculated results of PWR fuel isotopic composition by ORIGEN2 code

Suyama, Kenya; Murazaki, Minoru*; Mochizuki, Hiroki*; Nomura, Yasushi

JAERI-Tech 2001-074, 119 Pages, 2001/11


no abstracts in English

JAEA Reports

Preparation of data relevant to "equivalent uniform burnup" and "equivalent initial enrichment" for burnup credit evaluation

Nomura, Yasushi; Murazaki, Minoru*; Okuno, Hiroshi

JAERI-Data/Code 2001-029, 120 Pages, 2001/11


no abstracts in English

JAEA Reports

Study on safety of crystallization method applied to dissolver solution in fast breeder reactor reprocessing

; Fujine, Sachio; Asakura, Toshihide; Murazaki, Minoru*; *; *; *

JAERI-Research 99-027, 37 Pages, 1999/03


no abstracts in English

JAEA Reports

Base data for looking-up tables of calculation errors in JACS code system

Murazaki, Minoru*;

JAERI-Data/Code 99-019, 103 Pages, 1999/03


no abstracts in English

JAEA Reports

Three-dimensional thermofluid computer code CELVA-3D to evaluate the safety of hypothetical explosion in fuel reprocessing plants (Contract research)

Nishio, Gunji*; *; Kono, Koji*; *; Murazaki, Minoru*

JAERI-Data/Code 98-033, 235 Pages, 1998/11


no abstracts in English

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