Impact of nuclear data revised from JENDL-4.0 to JENDL-5 on PWR spent fuel nuclide composition

Watanabe, Tomoaki; Tada, Kenichi; Endo, Tomohiro*; Yamamoto, Akio*

Journal of Nuclear Science and Technology, 60(11), p.1386 - 1396, 2023/11

https://doi.org/10.1080/00223131.2023.2201603

The burnup calculations for estimating the nuclide composition of the spent fuel are highly dependent on nuclear data. Many nuclides in the latest version of the Japanese Evaluated Nuclear Data Library JENDL-5 were modified from JENDL-4.0 and the modification affects the burnup calculations. This study confirmed the validity of JENDL-5 in the burnup calculations. The PIE data of Takahama-3 was used for the validation. The effect of modifications of the parameters, e.g., cross sections and fission yields, from JENDL-4.0 to JENDL-5 on the nuclide compositions was quantitatively investigated. The calculation results showed that JENDL-5 has a similar performance to JENDL-4.0. The calculation results also revealed that the modifications of the cross sections of actinide nuclides, fission yields, and thermal scattering low data of hydrogen in HO affected the nuclide compositions of PWR spent fuels.

Report of summer holiday practical training 2020; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 3

Ishitsuka, Etsuo; Mitsui, Wataru*; Yamamoto, Yudai*; Nakagawa, Kyoichi*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Nagasumi, Satoru; Takamatsu, Kuniyoshi; Kenzhina, I.*; et al.

JAEA-Technology 2021-016, 16 Pages, 2021/09

JAEA-Technology-2021-016.pdf:1.8MB

As a summer holiday practical training 2020, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the downsizing of reactor core were studied by the MVP-BURN. As a result, it is clear that a 1.6 m radius reactor core, containing 54 (183 layers) fuel blocks with 20% enrichment of U, and BeO neutron reflector, could operate continuously for 30 years with thermal power of 5 MW. Number of fuel blocks of this compact core is 36% of the HTTR core. As a next step, the further downsizing of core by changing materials of the fuel block will be studied.

Preparation for restarting the high temperature engineering test reactor; Development of utility tool for auto seeking critical control rod position

Ho, H. Q.; Fujimoto, Nozomu*; Hamamoto, Shimpei; Nagasumi, Satoru; Goto, Minoru; Ishitsuka, Etsuo

Nuclear Engineering and Design, 377, p.111161_1 - 111161_9, 2021/06

https://doi.org/10.1016/j.nucengdes.2021.111161

Report of summer holiday practical training 2019; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 2

Ishitsuka, Etsuo; Nakashima, Koki*; Nakagawa, Naoki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Matsuura, Hideaki*; et al.

JAEA-Technology 2020-008, 16 Pages, 2020/08

JAEA-Technology-2020-008.pdf:2.98MB

As a summer holiday practical training 2019, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the U enrichment and burnable poison of the fuel, which enables continuous operation for 30 years with thermal power of 5 MW, were studied by the MVP-BURN. As a result, it is clear that a fuel with U enrichment of 12%, radius of burnable poison and natural boron concentration of 1.5 cm and 2wt% are required. As a next step, the downsizing of core will be studied.

SWAT4.0; The Integrated burnup code system driving continuous energy Monte Carlo codes MVP, MCNP and deterministic calculation code SRAC

Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki*

JAEA-Data/Code 2014-028, 152 Pages, 2015/03

JAEA-Data-Code-2014-028.pdf:13.39MB

There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0.

Effect of neutron induced reactions of neodymium-147 and 148 on burnup evaluation

Suyama, Kenya; Mochizuki, Hiroki*

Journal of Nuclear Science and Technology, 42(7), p.661 - 669, 2005/07

https://doi.org/10.3327/jnst.42.661

Burnup is important value for criticality safety evaluation of spent nuclear fuel. Nd-148 method is one of most important method to evaluate the burnup of post irradiation examination (PIE) samples, and well known that it has good accuracy. However, the evaluated burnup values could be perturbed by the neutron capture reaction of Nd-147 and Nd-148. And in the analysis of PIE data from PWR, the calculation results of Nd-148 have approximately more than 1% deviation from experiment. In this study, the contribution of neutron capture reaction of Nd-147 and Nd-148 to Nd-148 amount are discussed. Especially for Nd-147 contribution, it is shown that the current evaluated cross section of Nd-147 is not supported and the new evaluation is consistent with the analysis of PIE data. Possible perturbed amount of Nd-148 by both reactions is less than 0.7% for normal reactor operation condition, and it is approximately 0.1% for 30 GWd/t (BWR) and 40 GWd/t (PWR). Finally, we confirm again that Nd-148 method is good evaluation method.

Classification of criticality calculations with correlation coefficient method and its application to OECD/NEA burnup credit benchmarks phase III-A and II-A

Okuno, Hiroshi

Journal of Nuclear Science and Technology, 40(7), p.544 - 551, 2003/07

https://doi.org/10.3327/jnst.40.544

A method for classifying benchmark results of criticality calculations according to similarity was proposed in this paper. After formulation of the method utilizing correlation coefficients, it was applied to burnup credit criticality benchmarks Phase III-A and II-A, which were conducted by the Expert Group on Burnup Credit Criticality Safety under auspices of the OECD/NEA. Phase III-A benchmark was a series of criticality calculations for irradiated BWR fuel assemblies, whereas Phase II-A benchmark was a suite of criticality calculations for irradiated PWR fuel pins. These benchmark problems and their results were summarized. The correlation coefficients were calculated and sets of benchmark results were classified according to the criterion that the correlation coefficients were no less than 0.15 for Phase III-A and 0.10 for Phase II-A benchmarks. When a couple of results were in a same group, one result was found predictable from the other. An example was shown for each of the Benchmarks. The evaluated nuclear data seemed the main factor of errors.

Burnup importance function introduced to give an insight into the end effect

Okuno, Hiroshi; Sakai, Tomohiro*

Nuclear Technology, 140(3), p.255 - 265, 2002/12

https://doi.org/10.13182/NT02-A3337

In order to facilitate discussions based on quantitative analysis about the end effect, which is often talked about in connection to burnup credit in criticality safety evaluation of spent fuel, we introduced in this paper a burnup importance function. This function shows the burnup effect on the reactivity as a function of the fuel position; an explicit expression of this function was derived according to the perturbation theory. The burnup importance function was applied to the Phase IIA benchmark model that was adopted by the OECD/NEA Expert Group on Burnup Credit Criticality Safety. The function clearly displayed that burnup importance of the end regions increases (1) as burnup, (2) as cooling time, (3) in consideration of burnup profile, and (4) in consideration of fission products.

Burnup importance function and its application to OECD/NEA/BUC phase II-A and II-C models

Okuno, Hiroshi; Tonoike, Kotaro; Sakai, Tomohiro*

Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 8 Pages, 2002/10

As the burnup proceeds, reactivity of fuel assemblies for light water reactors decreases by depletion of fissile nuclides, especially in the axially central region. In order to describe the importance of the end regions to the reactivity change, a burnup importance function was introduced as a weighting function to a local burnup variation contributed to a reactivity decrease. The function was applied to the OECD/NEA/BUC Phase II-A model and a simplified Phase II-C model. The application to Phase II-A model clearly showed that burnup importance of the end regions increases as burnup and/or cooling time increases. Comparison of the burnup importance function for different initial enrichments was examined. The application result to the simplified Phase II-C model showed that the burnup importance function was helpful to find the most reactive fuel burnup distribution under the conditions that the average fuel burnup was kept constant and the variations in the fuel burnup were within the maximum and minimum measured values.

PLUTON: Three-group neutronic code for burnup analysis of isotope generation and depletion in highly irradiated LWR fuel rods

Lemehov, S.; Suzuki, Motoe

JAERI-Data/Code 2001-025, 338 Pages, 2001/08

JAERI-Data-Code-2001-025.pdf:26.87MB

PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO2, UO2-Gd2O3, inhomogeneous MOX, and UO2-ThO2. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of 92U233-239, 93Np237-239, 94Pu238-243, 95Am241-244 (including isomers), and 96Cm242-245. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained.