Validation of burnup calculation code SWAT4 by evaluation of isotopic composition data of mixed oxide fuel irradiated in pressurized water reactor

加圧水型原子炉の使用済みMOX燃料の同位体組成評価による燃焼計算コードの検証

鹿島 陽夫; 須山 賢也; 望月 弘樹*

Kashima, Takao; Suyama, Kenya; Mochizuki, Hiroki*

福島第一原子力発電所事故により日本では核燃料サイクル計画が停滞しているが、過剰なプルトニウムを減らすためにMOX燃料として使用する必要がある。また、福島第一原子力発電所ではMOX燃料を使用していたということからも、MOX燃料の燃焼解析の精度は重要である。したがって、計算コードとライブラリの精度評価が必要である。そこで、使用済燃料組成測定(ARIANE: Actinides Research in a Nuclear Elements)のデータを解析することにより、統合化燃焼計算コードシステム(SWAT4: Step Wise Burnup Analysis Code System 4)の検証を行った。解析の結果、主要なウラン,プルトニウム同位体の実験値と計算値の差は5%以下であった。これは、ウラン燃料の場合と同程度の精度であり、SWAT4はMOX燃料の燃焼解析に対して十分な精度を持っていることが分かった。

The nuclear fuel cycle program of Japan would be delayed because of the impact of the Fukushima Daiichi NPP accident in 2011. Excessive plutonium, however, has to be utilized as mixed-oxide (MOX) fuel to reduce the quantity of plutonium possessed by Japan. Calculation codes and libraries adopted in the fuel cycle analyses of MOX fuel should be benchmarked based on comparison between calculation results and experimental data. From another viewpoint, nuclide inventory analyses of MOX fuel is important for evaluations of the Fukushima accident because MOX fuel has been loaded in the Unit 3 reactor. ARIANE is a PIE program which includes measurements of nuclide compositions of spent MOX fuels discharged from both of pressurized and boiling water reactors. In this study, the PIE data of MOX fuels irradiated in a pressurized water reactor were analyzed by the integrated burnup code system SWAT4 that combines the point burnup system ORIGEN2 and neutron transport calculation solvers, the continuous energy Monte Carlo code MVP or MCNP, and the deterministic neutronics calculation code SRAC. The calculation results of SWAT4 have generally same trends with the case of UO fuel analyses. For major uranium and plutonium isotopes, deviations less than 5% were obtained. This means that SWAT4 has the same accuracy to predict isotopic compositions of irradiated MOX fuel with the case of UO fuel. The radial distribution of isotopes in a pellet was also analyzed, whose results were compared with that measured by SIMS. SWAT4 predicted well the isotope and burnup distributions in an irradiated MOX pellet.