Keisuke Okumura

Particle and Heavy Ion Transport code System, PHITS, version 2.52

An upgraded version of the Particle and Heavy Ion Transport code System, PHITS2.52, was developed and released to the public. The new version has been greatly improved from the previously released version, PHITS2.24, in terms of not only the code itself but also the contents of its package, such as the attached data libraries. In the new version, a higher accuracy of simulation was achieved by implementing several latest nuclear reaction models. The reliability of the simulation was improved by modifying both the algorithms for the electron-, positron-, and photon-transport simulations and the procedure for calculating the statistical uncertainties of the tally results. Estimation of the time evolution of radioactivity became feasible by incorporating the activation calculation program DCHAIN-SP into the new package. The efficiency of the simulation was also improved as a result of the implementation of shared-memory parallelization and the optimization of several time-consuming algorithms. Furthermore, a number of new user-support tools and functions that help users to intuitively and effectively perform PHITS simulations were developed and incorporated. Due to these improvements, PHITS is now a more powerful tool for particle transport simulation applicable to various research and development fields, such as nuclear technology, accelerator design, medical physics, and cosmic-ray research.

JENDL-4.0 Benchmarking for Fission Reactor Applications

Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0.

Determination of 79Se and 135Cs in Spent Nuclear Fuel for Inventory Estimation of High-Level Radioactive Wastes

79Se and 135Cs are long-lived fission products and are found in high-level radioactive waste (HLW). The estimation of their inventories in HLW is essential for the safety assessment of geological disposal, owing to their mobility in the strata. In this study, the amounts of 79Se and 135Cs in a spent nuclear fuel solution were measured. About 5 g of irradiated UO2 fuel discharged from a commercial Japanese pressurized water reactor (PWR) with a burn-up of 44.9 GWd/t was sampled and dissolved with 50mL of 4M nitric acid in a hot cell for 2 h. After Se and Cs were chemically separated, the amounts of 79Se and 135Cs in the spent nuclear fuel solution were measured by inductively coupled plasma quadrupole mass spectrometry (ICP-QMS). The amounts of 79Se and 135Cs were 5:2 ± 1:5 and 447 ± 40 g/MTU, respectively. The results presented in this study, which are the first postirradiation experimental data in Japan, showed good agreement with those obtained by the ORIGEN2 code using the data library of JENDL-3.3.

JENDL Actinoid File 2008

JENDL Actinoid File 2008 (JENDL/AC-2008) was released in March 2008. It includes nuclear data for neutron-induced reactions for 79 nuclides from Ac (Z = 89) to Fm (Z = 100). The neutron energy range is 10−5 eV to 20 MeV. Almost alldata for 62 actinoids in JENDL-3.3 were revised. New evaluations were performed for 17 nuclides, which have half-lives longer than one day. A new comprehensive theoretical model code CCONE was widely used for the evaluation of cross sections and neutron emission spectra. Thermal cross sections for many nuclides were revised based on experimental data. Resonance parameters were readjusted to reproduce them. Simultaneous evaluations of fission cross sections were performed for six important nuclei. The least-squares fitting code GMA was used for the evaluation of fission cross sections for minor actinoids. In this paper, we present the evaluation methods and results of the JENDL/AC-2008.

Sensitivity Analysis of Fission Product Concentrations for Light Water Reactor Burned Fuel

The accurate prediction of fission product concentrations (FPCs) is necessary for application of the burnup credit to nuclear facilities. In order to specify important nuclear data for the accurate prediction of FPC, we extensively evaluate the sensitivities of FPC to nuclear data with the depletion perturbation theory. The target fission products are twelve important ones for the burnup credit, Mo-95, Tc-99, Rh-103, Nd-143, Nd-145, Sm-147, Sm-149, Sm-150, Sm-152, Cs-133, Eu-153, and Gd-155. The present study successfully specifies the important nuclear data both in a UO2 cell and in a MOX cell. While the obtained sensitivities are mostly similar to each other between the UO2 and MOX cells, large differences are observed in some cases, such as the Gd-155 concentration. It is clearly shown that such differences between the UO2 and MOX cells come from differences in cumulative fission yields between U-235 and Pu-239 and differences in neutron flux energy spectra.

Isotope dilution inductively coupled plasma mass spectrometry for determination of 126Sn content in spent nuclear fuel sample

The 126Sn content in a spent nuclear fuel solution was determined by isotope dilution inductively coupled plasma mass spectrometry (ID-ICP-MS) for its inventory estimation in high-level radioactive waste. A well-characterized irradiated UO2 fuel sample dissolved in a hot cell was used as a sample to evaluate the reliability of the methodology. Prior to the ICP-MS measurement, Sn was separated from Te (126Te), which causes major isobaric interference in the determination of 126Sn content, along with highly radioactive coexisting elements, such as Sr (90Sr), Y (90Y), Cs (137Cs) and Ba (137m Ba), using an anion-exchange column. The absence of counts attributed to Te at m/z = 125, 128, and 130 in the Sn-containing effluent (Sn fraction) indicates that Te was completely removed from the anion-exchange column. After washing, Sn retained on the column was readily eluted with 1 M HNO3 accompanied with approximately 80% of the Cd and 0.03% of the U in the initial sample. Owing to the presences of Cd and U in Sn fraction, the measurements of 116Sn and 119Sn were affected by the isobaric 116Cd and the doubly charged 238U2+ion, resulting in the positive bias of the determined values. With the exception of the isotopic ratios including 116Sn and 119Sn, 117Sn/126Sn, 118Sn/126Sn, 120Sn/126Sn, 122Sn/126Sn and 124Sn/126Sn were successfully determined and showed good agreement with those obtained through ORIGEN2 calculations. The measured concentration of 126Sn in the spent nuclear fuel sample solution was 0.74 ± 0.14 ng/g, which corresponds to 23.0 ± 4.5 ng per gram of the irradiated UO2 fuel (excluding the presence of 126Sn in the insoluble residue). The results reported in this paper are the first experimental values of 126Sn content and isotope ratios in the spent nuclear fuel solution originating in spent nuclear fuel irradiated at a nuclear power plant in Japan.

A study of the generation of 232U in UO2 and MOX fuels

To clarify the generation pathway of 232U, an important nuclide for dose evaluation at various stages in the reuse of uranium, concentrations of 232U generated through various pathways were evaluated for UO2 and mixed oxide (MOX) fuels. Burnup calculation was conducted with ORIGEN2.2 code adopting ORLIBJ40 library, a set of cross-section libraries based on JENDL-4.0. It was found that differences in 232U concentrations in UO2 and MOX fuels mainly arise from differences in the initial compositions of 234U, 235U, and 236U. It was also found that the contribution of plutonium and americium isotopes in MOX fuels is small compared with that of uranium isotopes. The results clarified that the capture cross sections of 230Th, 231Pa, 235U, and 236U, as well as the (n,2n) cross sections of 237Np and 238U, have a large effect on the generation of 232U. Additional investigation showed that 232U concentration is strongly affected not only by time after irradiation but also by time before irradiation.

Nuclear Reactor Calculations

The most fundamental evaluation quantity of the nuclear design calculation is the effective multiplication factor (k eff ) and neutron flux distribution. The excess reactivity, control rod worth, reactivity coefficient, power distribution, etc. are undoubtedly inseparable from the nuclear design calculation. Some quantities among them can be derived by secondary calculations from the effective multiplication factor or neutron flux distribution. Section 2.1 treats the theory and mechanism to calculate the effective multiplication factor and neutron flux distribution in calculation programs (called codes). It is written by Keisuke Okumura.

Comparison of Post-Irradiation Experimental Data and Theoretical Calculations for Inventory Estimation of Long-Lived Fission Products in Spent Nuclear Fuel

The inventory estimation of long-lived fission products (LLFP) is essential for the long-term safety assessment of a geological disposal of high-level radioactive waste (HLW). 79 Se and 135 Cs are a main contributor to the total dose from the geological repository of HLW, owing to their solubility in the strata. In this study, the post-irradiation experimental data of LLFPs, such as 79 Se, 99 Tc, 126 Sn and 135 Cs, were compared with ORIGEN2 calculation using the data library of JENDL-3.3. A fragment of the UO2 fuel pellet irradiated in a commercial Japanese PWR was dissolved with nitric acid in a hot cell. The resultant solution was filtered to remove insoluble residue. After Se, Tc, Sn, and Cs were chemically separated, the concentrations of 79 Se, 99 Tc, 126 Sn and 135 Cs were determined with an inductively coupled plasma quadrupole mass spectrometer (ICP-QMS). The concentration of 79 Se, 99 Tc, 126 Sn and 135 Cs in the sample solution were 0.78 ± 0.22, 101 ± 24, 3.2 ± 0.6, and 68 ± 6.0 ng/g of the sample solution (ng/g-sol.), respectively. The results for 79 Se and 135 Cs obtained in this study showed good agreement with those obtained through ORIGEN2 calculation. This indicates that ORIGEN2 calculation is applicable to the estimation of the amounts of 79 Se and 135 Cs generated during irradiation. In contrast, the experimentally determined concentration of 99 Tc and 126 Sn were equivalent to approximately 70% and 60%, respectively, of those obtained through ORIGEN2 calculation.Copyright

On the design of a remotely-deployed detection system for reactor assessment at Fukushima Daiichi

The premise behind this research is the design of a system that will allow fuel debris characterisation at Fukushima Daiichi. The precise location of the debris is not known for example as to whether it remains within the reactor pressure vessel or it has leaked through into the base of the pedestal below, additionally the state of the fuel is also in question as to whether this has corroded from within its encasing or if it is intact. The most likely scenario is a combination of all four of these situations. The flooding of the reactor floors immediately following the Fukushima accident adds an extra element of complexity for the detection system requiring it to be submersible and to hold any detector system in water tight confinement. The research carried out has involved extensive modifications to a previously-designed low-cost small-scale AVEXIS submersible inspection vehicle and the incorporation of a variety of radiation detectors. The latter has been designed to allow for mapping and determination of the situation that is present within the primary containment vessels. The challenges addressed with the detection system arise from the high dose rates that have been recorded around the reactor pressure vessels which can be as high as 1000 Gy/hr. In such a harsh environment not only will the radiation detectors struggle to operate but the components that make up the remote-operated vehicle are also likely to suffer radiation damage after only a relatively short period of time. The research presented here evaluates the components currently incorporated into the AVEXIS system in terms of their radiation tolerability as well as presenting the combination of detectors to be used in the remote probe for the investigation of the fuel debris.