Calculations for radioactivity evaluation of research reactors for near surface disposal and their application methods

Kochiyama, Mami

Kaku Deta Nyusu (Internet), (133), p.76 - 81, 2022/10

http://www.aesj.or.jp/~ndd/ndnews/No133.html

The outline of the presentation at the joint session of Research Committee for Nuclear Data and Subcommittee on Nuclear Data in the Atomic Energy Society of Japan 2022 Autumn Meeting was contributed to Nuclear Data News. As part of the study on the near surface disposal of waste from research facilities, we are studying a method for evaluating the radioactivity inventory of waste generated by the dismantling of research reactors. In the radioactivity evaluation of the research reactor, we have investigated the method of calculating the neutron transport in the reactor and using the obtained neutron spectrum to calculate the activation of the internal structure by the ORIGEN-S code. In recent years, we have introduced and evaluated libraries created based on JENDL-4.0 and JENDL/AD-2017, and we will introduce the status of their examination. And we will introduce how to apply the results obtained by the radioactivity evaluation calculation to burial disposal.

Study on radioactivity evaluation method of research reactors using DORT and MCNP codes

Kochiyama, Mami; Sakai, Akihiro

JAEA-Technology 2022-009, 56 Pages, 2022/06

JAEA-Technology-2022-009.pdf:4.15MB

It is necessary to evaluate radioactivity inventory in wastes before disposal of low-level radioactive wastes generated from dismantling research reactors. It is efficient for owners of each research reactor to use a common radioactive evaluation method in order to comply with the license application for disposal facility. In this report, neutron transport and activation calculations were carried out for the Rikkyo University research reactor in order to examine a common radioactivity evaluation method for burial disposal of radioactive wastes generated by dismantling. We adopted the neutron transport codes DORT and MCNP and the activation code ORIGEN-S with cross-section libraries based on JENDL-4.0 and JENDL/AD-2017. The radioactivity concentrations obtained by the radiochemical analysis and both calculation codes were in agreement by 0.4 to 3 times. Therefore, by appropriately considering this difference, the radioactivity evaluation method by DORT, MCNP and ORIGEN-S can be applied to the radioactivity evaluation for buried disposal. In order to classify wastes from dismantling by clearance or buried disposal method according to their radioactivity levels, we also created radioactivity concentration distributions in the concrete area and graphite thermal column area.

Development of burnup/depletion calculation code based on ORIGEN2 cross-section libraries and Chebyshev rational approximation method, CRAMO

Yokoyama, Kenji; Jin, Tomoyuki*

JAEA-Data/Code 2021-001, 47 Pages, 2021/03

JAEA-Data-Code-2021-001.pdf:1.85MB

A new burnup/depletion calculation code, CRAMO, was developed by combining an ORIGEN2 cross-section library set, ORLIB, based on Japanese evaluated nuclear data library, JENDL, and a burnup/depletion solver based on Chebyshev rational approximation method. CRAMO uses the ORIGEN2 cross-section library set ORLIBJ40 based on JENDL-4.0, and the burnup/depletion solver implemented in the versatile reactor analysis code system, MARBLE. It was confirmed that results of CRAMO agreed well with those of ORIGEN2 for burnup/depletion and radioactivity calculation cases. The development of CRAMO made it possible to use ORLIB without using ORIGEN2. It will be possible to provide an easy-to-use processed JENDL data set for burnup/depletion and radioactivity calculations in combination with a burnup/depletion based on Chebyshev rational approximation method. The present version of CRAMO is a subset of ORIGEN2 and can compute only compositions and radioactivities after irradiation. However, since various kinds of outputs of ORIGEN2 can be evaluated by using the composition, it is possible to reproduce many functions of ORIGEN2 by adding post-processing modules.

Feasibility study on tritium recoil barrier for neutron reflectors of research and test reactors

Kenzhina, I.*; Ishitsuka, Etsuo; Ho, H. Q.; Sakamoto, Naoki*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*

Fusion Engineering and Design, 164, p.112181_1 - 112181_5, 2021/03

https://doi.org/10.1016/j.fusengdes.2020.112181

Tritium release into the primary coolant during operation of the JMTR (Japan Materials Testing Reactor) and the JRR-3M (Japan Research Reactor-3M) had been studied. It is found that the recoil release by Li(n,)H reaction, which comes from a chain reaction of beryllium neutron reflectors, is dominant. To prevent tritium recoil release, the surface area of beryllium neutron reflectors needs to be minimum in the core design and/or be shielded with other material. In this paper, as the feasibility study of the tritium recoil barrier for the beryllium neutron reflectors, various materials such as Al, Ti, V, Ni, and Zr were evaluated from the viewpoint of the thickness of barriers, activities after long-term operations, and effects on the reactivities. From the results of evaluations, Al would be a suitable candidate as the tritium recoil barrier for the beryllium neutron reflectors.

Development of inventory calculation modules using ORIGEN-S for decommissioning

Matsuda, Norihiro; Konno, Chikara; Ikehara, Tadashi; Okumura, Keisuke; Suyama, Kenya*

JAEA-Data/Code 2020-003, 33 Pages, 2020/03

JAEA-Data-Code-2020-003.pdf:1.85MB

Data handling modules for the radioactivity calculation code, ORIGEN-S, are developed for the reliable evaluations of radioactivity inventory. By using these modules, an activation cross-section data library for the ORIGEN-S code is updated easily and effectively based on a facility-specific neutron spectrum and multi-group neutron activation cross-section library for decommissioning of nuclear facilities, MAXS2015. In order to guarantee the reliability of the radioactivity calculations, functions of data verification in a visual way and numerical comparison between before and after the data processing are also prepared.

Burn-up characteristics and criticality effect of impurities in the graphite structure of a commercial-scale prismatic HTGR

Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo

Nuclear Engineering and Design, 326, p.108 - 113, 2018/01

AA2015-0964.pdf:0.64MB

https://doi.org/10.1016/j.nucengdes.2017.11.003

Burn-up characteristics and criticality of impurity contained into graphite structure for commercial scale prismatic High Temperature Gas-cooled Reactor (HTGR) have been investigated. For HTGR, of which the core is filled graphite structure, the impurity contained into the graphite has unignorable poison effect for criticality. Then, GTHTR300, commercial scale HTGR, employed high grade graphite material named IG-110 to take into account the criticality effect for the reflector blocks next to fuel blocks. The fuel blocks, which should also employ IG-110, employ lower grade graphite material named IG-11 from the economic perspective. In this study, the necessity of high grade graphite material for commercial scale HTGR is reconsidered by evaluating the burn-up characteristics and criticality of the impurity. The poison effect of the impurity, which is used to be expressed by a boron equivalent, reduces exponentially like burn-up of B, and saturate at a level of 1 % of the initial value of boron equivalent. On the other hand, the criticality effect of the boron equivalent of 0.03 ppm, which corresponds to a level of 1 % of IG-11 shows ignorable values lower than 0.01 %k/kk' for both of fuel blocks and reflector blocks. The impurity can be represented by natural boron without problem. Therefore, the poison effect of the impurity is evaluated with whole core burn-up calculations. As a result, it is concluded that the impurity is not problematic from the viewpoint of criticality for commercial scale HTGR because it is burned clearly until End of Cycle (EOC) even with the low grade graphite material of IG-11. According to this result, more economic electricity generation with HTGR is expected by abolishing the utilization of IG-110.

Calculation of decay heat by new ORIGEN libraries for high temperature engineering test reactor

Simanullang, I. L.*; Honda, Yuki; Fukaya, Yuji; Goto, Minoru; Shimazaki, Yosuke; Fujimoto, Nozomu*; Takada, Shoji

JAEA-Technology 2015-032, 26 Pages, 2016/01

JAEA-Technology-2015-032.pdf:2.07MB

Decay heat of the High Temperature Engineering Test Reactor had been evaluated by the Shure Equation and/or ORIGEN code based on the LWR's data. However, to evaluate more accurately, a suitable method must be considered because of the differences neutron spectrums from the LWRs. Therefore, the decay heat and the generated nuclides for the neutron spectrums of the core with different graphite moderator amount were calculated by the ORIGEN2 code. As a result, it is clear that the calculated decay heats are similar value with LWRs for about one year after the reactor shutdown, and that the significant differences are observed on the longer period affected by the generated nuclides such as Y, Cs, Pr, Rh, Am etc. It is also clear that the dose is affected by Pu on the initial stage after the reactor shutdown.

Study on the prediction accuracy of nuclide generation and depletion with JENDL

Okumura, Keisuke; Oki, Shigeo*; Yamamoto, Munenari*; Matsumoto, Hideki*; Ando, Yoshihira*; Tsujimoto, Kazufumi; Sasahara, Akihiro*; Katakura, Junichi; Matsumura, Tetsuo*; Aoyama, Takafumi*; et al.

JAERI-Research 2004-025, 154 Pages, 2005/01

JAERI-Research-2004-025.pdf:19.46MB

This report summarizes the activity (FY2000-2003) of Working Group (WG) on Evaluation of Nuclide Generation and Depletion under Subcommittee on Nuclear Fuel Cycle of Japanese Nuclear Data Committee. In the WG, analyses of Post Irradiation Examinations have been carried out for UO and MOX fuels irradiated in PWRs, BWRs and FBRs, and for actinide samples irradiated in fast reactors, by using ORIGEN or more detailed calculation codes with their libraries based on JENDL-3.2, JENDL-3.3 and other foreign nuclear data files. From these results, current prediction accuracy and problems for evaluation of nuclide generation and depletion are discussed. Furthermore, this report covers other products of our activity; development of the ORIGEN libraries for PWR, BWR and FBR based on JENDL-3.3, study on introduction of neutron spectrum index to ORIGEN calculations, and results of questionnaire survey on desirable accuracy of ORIGEN calculations.

Validation of integrated burnup code system SWAT2 by the analyses of isotopic composition of spent nuclear fuel

Suyama, Kenya; Mochizuki, Hiroki*; Okuno, Hiroshi; Miyoshi, Yoshinori

Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 10 Pages, 2004/04

This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models.

Validation of minor actinide cross sections by studying samples irradiated for 492 days at the dounreay prototype fast reactor, 2; Burnup calculations

Tsujimoto, Kazufumi; Kono, Nobuaki; Shinohara, Nobuo; Sakurai, Takeshi; Nakahara, Yoshinori; Mukaiyama, Takehiko; Raman, S.*

Nuclear Science and Engineering, 144(2), p.129 - 141, 2003/06

https://doi.org/10.13182/NSE03-A2348

To evaluate neutron cross-section data of minor actinides, separated actinide samples and dosimetry samples were irradiated at the Dounreay Prototype Fast Reactor for 492 effective full power days. Based on the burnup calculations of major actinide and dosimetry samples, the neutron flux distribution and the flux level were adjusted at the locations where minor actinide samples were irradiated. The burnup calculations were carried out for minor actinides using the determined flux distribution and flux level. This paper discusses the burnup calculations and the validation of minor actinide cross-section data in evaluated nuclear data libraries. We find that we can obtain reliable FIMA (fission per initial metallic atom) values by using the Nd method except that the uncertainties in the FIMA values are large for U, Pu, Am isotopes, and Cm isotopes because the Nd yields are known poorly for these isotopes and are probably overestimated. For these isotopes, measurements to improve the fission-yield data are needed. We also find that, in general, the JENDL-3.2 nuclear data for the minor actinides are adequate for the conceptual design study of transmutation systems. But, there are some nuclides (especially Pu and Pu) for which new measurements are needed particulary if the minor actinides constitute a major part of the nuclear fuel.