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

https://doi.org/10.1016/j.anucene.2011.01.025

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.

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

JAEA-Technology-2009-022.pdf:5.36MB

We have re-evaluated the dose measurements at SILENE with TLDs for neutrons and with TLDs for 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 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 rays.

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

https://doi.org/10.3327/jnst.46.193

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 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.

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

https://doi.org/10.1016/j.anucene.2007.02.002

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.

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

JAERI-Tech-2003-029.pdf:7.04MB

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 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, , was evaluated from the measurement data of pulsed neutron source experiments. We also calculated distribution of neutron flux and 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.

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

JAERI-Data-Code-2001-029.pdf:6.16MB

no abstracts in English

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

Murazaki, Minoru*;

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

JAERI-Data-Code-99-019.pdf:3.69MB

no abstracts in English