Satoshi Shimakawa

Impact of Revised Thermal Neutron Capture Cross Section of Carbon Stored in JENDL-4.0 on HTTR Criticality Calculation

Recently, high-temperature gas-cooled reactors (HTGRs) have been receiving particular attention as one of the Generation IV nuclear reactor systems in the world, because of its excellence in safety, economical efficiency, and nuclear proliferation resistance, and applicability of nuclear power as a heat source for the thermochemical Iodine-Sulfur (IS) process, by which hydrogen is produced without the release of carbon dioxide. Many countries, then, have been performing design studies on HTGRs. Meanwhile, in Japan, the Japan Atomic Energy Agency (JAEA) has been conducting design studies on commercial HTGRs. JAEA has also been operating the High-Temperature Engineering Test Reactor (HTTR), which is a testing HTGR, and it has yielded useful data for conducting design studies on commercial HTGRs. The improvement of accuracy of the HTGR neutronics calculations allows the design of commercial HTGRs for low cost and high performance. Generally, the accuracy of neutronics calculation results depends on the nuclear data library used in the calculations; thus, the evaluation of the applicability of nuclear data libraries to HTGR neutronics calculations is one of the important subjects. In the past, the neutronics calculations for the HTTR critical approach were performed with the three major nuclear data libraries, namely, JENDL-3.3 (Japan), ENDF/BVI.5 (U.S.A), and JEFF-3.0 (Europe). As a result, JENDL3.3 yielded keff values that were in better agreement with the experimental results than the other libraries. Additionally, it was found that the discrepancies of the keff values between JENDL-3.3 and the other libraries are mainly caused by the slight difference in the neutron capture cross section of carbon at 0.0253 eV among the libraries, and we focused on the accuracy of this cross section as one of the important subjects for the improvement of the neutronics calculations for the HTGRs. JENDL-3.3 showed better applicability to the HTTR criticality calculations than the other libraries as mentioned above, but still overestimated the keff values by 0.5– 1.1% k. By overestimating the keff values, the calculation result of the loaded number of fuel columns achieving the first criticality did not agree with the experimental results. These problems were not resolved until today, despite our refinement efforts, such as the description of the core geometry and the concentration of the components. Meanwhile, the neutron capture cross section of carbon at 0.0253 eV stored in each nuclear data library had not been revised for a long time. Thus, we proposed that this cross section should be revised based on the latest measurement data, and also predicted that the problem of overestimating the keff values will be resolved by revising the cross section to be about 10% larger than that of JENDL-3.3. In May 2010, the latest JENDL, JENDL-4.0, was released by JAEA. In JENDL-4.0, our proposal with the prediction was applied, and the neutron capture cross section of carbon at 0.0253 eV was revised based on the latest measurement data. Accordingly, the problem of overestimating the keff values in the HTTR criticality calculations was expected to be addressed. This paper describes the investigation of the applicability of JENDL-4.0 to the HTTR criticality calculations.

Neutronics calculations of HTTR with several nuclear data libraries

Benchmark calculations for several HTTR core states were performed with four cross-section sets which were generated from JENDL-3.3, JENDL-3.2, ENDF/B-VI.8 and JEFF-3.0 using a continuous energy Monte Carlo code MVP. The core states were a critical approach in which an annular core was formed at room temperature and solid cores at room temperature and at full power operation. Study of keff discrepancies caused by difference of the nuclear data libraries and identification of nuclides which have large effects on the keff discrepancies were carried out. Comparison of the respective keff from calculations and experiments was also carried out. As the results, for each of the HTTR core states, JENDL-3.3 yields a keff agreeing with the experiments within 1.5%Δk, JENDL-3.2 yields keff agreement within 1.7%Δk, and ENDF/B-VI.8 and JEFF-3.0 yield keff agreement within 1.8%Δk. There is little keff discrepancy between ENDF/B-VI.8 and JEFF-3.0. The keff between JENDL-3.3 and JENDL-3.2 is caused by difference of 235U data and has temperature dependency. The keff discrepancy between JENDL-3.3 and ENDF/B-VI.8 or JEFF-3.0 is mainly caused by difference in graphite data.

Study of Tritium Production for Fusion Reactors Using High-Temperature Gas-Cooled Reactors

The performance of a high-temperature gas-cooled reactor as a tritium production device was examined. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) was assumed as the calculation target of a typical gas-cooled reactor, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the entire-core region of GTHTR300 were carried out considering its unique double heterogeneity structure. It was shown that gas-cooled reactors with thermal output power of 3 GW in all can produce 6~10 kg of tritium in a year.

Core Design Study of Small-Sized High Temperature Reactor for Electricity Generation

A High Temperature Gas-cooled Reactor (HTGR) has several features different from conventional light water reactors such as inherent safety characteristics, high thermal efficiency and high economy. On the other hand, one of disadvantages of the HTGR with a prismatic core is to require rather long-term and expensive refueling, resulting in relatively long maintenance period and high cost. To solve the disadvantage, the present study challenges the core design of a small-sized reactor for long refueling interval by increasing core size, fuel loading and fuel burn up compared with the High Temperature engineering Test Reactor (HTTR). The preliminary burn-up calculation suggested that approximately 6 years of long refueling interval was found to be reasonably achieved. A refueling interval longer than 6 years may be possible by decreasing further power density, subsequently larger core size with operational reactor power of 120MWt, but this idea was not taken by the requirement of the reactor that the core size shall be accommodated reasonably in the core with double size of the HTTR at maximum.Copyright

Neutron and thermal calculation of blanket for high power operating condition of fusion reactor

Abstract As the international thermonuclear experimental reactor (ITER) project has proceeded to the engineering design activity (EDA) phase, the design parameters of the reactor have been reconsidered from those in the conceptual design activity (CDA) phase. The fusion power is one of the parameters being reconsidered. In the primitive assessment, the possibility of a fusion power of 1.5−3.0 GW was indicated. With simultaneous enlargement of the reactor dimensions, the neutron wall loading would be 1.2−2.3 times as high as that in the CDA phase. A tritium breeding blanket with ceramic breeder pebbles, proposed by Japan in the CDA phase, has been investigated for the accommodation of these high fusion powers. For a layered pebble bed-type blanket, the breeder and neutron multiplier layer thicknesses have been adjusted to allow a power variation from 1.5 to 3.0 GW. The neutron and thermal performance of the blanket design have been evaluated, and the tritium inventory in the blanket has been estimated. It has been shown that the layered pebble bed blanket is basically applicable to the high power operating conditions.