Evaluation of sodium radioactivity in the primary system of the prototype fast reactor Monju

Mori, Tetsuya; Ohgama, Kazuya; Hazama, Taira

Nuclear Technology, 209(7), p.1008 - 1023, 2023/07

https://doi.org/10.1080/00295450.2023.2181044

In this study, the sodium radioactivity of Na and Na in the primary system measured in the prototype fast breeder reactor Monju was evaluated, and the reliability of measurements and calculations was examined. The calculated-to-experiment (C/E) values and their uncertainties for Na and Na radioactivities were 0.97-1.07 and 8.1%-11.0% and 1.03-1.16 and 23.3%-24.1%, respectively, using JENDL-4.0 nuclear data library. The Na radioactivity calculated with ENDF/B-VIII.0 was larger by 40% than those calculated with JENDL-4.0 and JEFF-3.3 due to the Na(n,2n) cross-section discrepancy. The importance of the Na neutron capture effect was also confirmed herein for the accurate evaluation of the Na radioactivity. The experimental data was judged to be useful for validating the calculation method for improving the reliability of the future designs of sodium-cooled fast reactors.

Chapter 3, Prototype reactor Monju

Hazama, Taira

Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, p.87 - 161, 2022/07

https://doi.org/10.1016/B978-0-12-824076-2.00003-1

Evaluation of fixed absorber reactivity measurement in the prototype fast reactor Monju

Ohgama, Kazuya; Katagiri, Hiroki; Takegoshi, Atsushi*; Hazama, Taira

Nuclear Technology, 207(12), p.1810 - 1820, 2021/12

https://doi.org/10.1080/00295450.2020.1846481

Uncertainty evaluation of anticipated transient without scram plant response in the Monju reactor considering reactivity coefficients within the design range

Sotsu, Masutake; Hazama, Taira

Journal of Energy and Power Engineering, 13(11), p.393 - 403, 2019/11

https://doi.org/10.17265/1934-8975/2019.11.001

This paper describes methods and results of the uncertainty evaluation of the significant plant response analysis of the reactor trip failure event, i.e. anticipated transient without scram of the Japanese prototype fast breeder reactor Monju. Unprotected loss of heat sink has a relatively large contribution to the core damage frequency due to reactor trip failure. The uncertainty of the allowable time to core damage in this event by plant transient response analysis, so far, has been estimated with considering the range of reactivity coefficients. There are some cases where core damage is considered to be avoided. Specifically, it is assumed that the core damage due to the ULOHS event would be avoided if the sodium temperature at the pump inlet stays below 650C for 1 h; otherwise the possibility of cavitation occurring in the hydrostatic bearing increases. In this study, a method is developed to search for a solution as an inverse problem of multiple input variables that satisfy the temperature condition. This paper, as a first step, describes input conditions and probability to satisfy the temperature are evaluated through analyses treating input parameters, reactivity coefficients and kinetic parameters, as variables within the design range.

Irradiation induced reactivity in Monju zero power operation

Takano, Kazuya; Maruyama, Shuhei; Hazama, Taira; Usami, Shin

Proceedings of Reactor Physics Paving the Way Towards More Efficient Systems (PHYSOR 2018) (USB Flash Drive), p.1725 - 1735, 2018/04

Irradiation dependence of the core excess reactivity was investigated for the Monju system startup tests at zero-power carried out in 2010. The excess reactivity basically decreases with the decay of Pu in zero-power operation. However, the excess reactivity little changed in the two month period of the startup tests, which suggests a positive reactivity insertion during the period. The investigated irradiation dependence shows that the positive reactivity increases with reactor operation and mostly saturates by the fission-dose attained during the Monju zero-power operation in a month (10 fissions/cm). The saturated positive reactivity is equivalent to approximately 47% of the initially accumulated self-irradiation damage recovery assuming the defects were recovered by the fission-fragment irradiation in the reactor operation.

Development of the versatile reactor analysis code system, MARBLE2

Yokoyama, Kenji; Jin, Tomoyuki; Hirai, Yasushi*; Hazama, Taira

JAEA-Data/Code 2015-009, 120 Pages, 2015/07

JAEA-Data-Code-2015-009.pdf:1.93MB

The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added inMARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, -ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability.

Basic experiment for kinetics analysis in sub-critical sate

Kitano, Akihiro; Kishimoto, Yasufumi; Misawa, Tsuyoshi*; Hazama, Taira

KURRI Progress Report 2013, 1 Pages, 2014/10

http://www.rri.kyoto-u.ac.jp/PUB/report/PR/ProgRep2013/CO3.pdf#page=6

The approach to criticality is conventionally performed by the inverse multiplication method. The method uses neutron count rate at a steady state attained in a certain waiting time after a reactivity insertion; thus it requires long time (for example, several hours from the startup in Monju reactor). We have developed a more efficient method based on Critical Index (CI) featuring the time behavior of delayed neutrons.

Method development and reactor physics data evaluation for improving prediction accuracy of fast reactors' minor actinides transmutation performance

Takeda, Toshikazu*; Hazama, Taira; Fujimura, Koji*; Sawada, Shusaku*

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 15 Pages, 2014/09

A national project started in 2013 in Japan entitled "technology development for the environmental burden reduction". The present study is one of the studies adopted as the national project. We are aiming to develop MA transmutation core concepts harmonizing MA transmutation performance with core safety and to improve design accuracy related to MA transmutation performance. To validate and improve design accuracy of the high safety and high MA transmutation performance of SFR cores, we develop methods for calculating the transmutation rate of individual MA nuclides and estimating uncertainty of MA transmutation by using burnup sensitivity. Also we develop reliable reactor physics database to reduce the uncertainty of MA transmutation calculations. The overall consistency of the measured data is investigated by evaluating the usefulness of conventional static data as well as those related to MA transmutation obtained from various facilities like Monju, Joyo, FCA, BFS and PFR.

Development of versatile nuclear heating calculation system in MARBLE

Hazama, Taira; Yokoyama, Kenji

Journal of Nuclear Science and Technology, 50(5), p.525 - 533, 2013/05

https://doi.org/10.1080/00223131.2013.785280

A versatile nuclear heating calculation system is developed in MARBLE, a neutronics calculation platform based on the object-oriented language. The system provides a variety of heating calculation methods from the conventional simple methods based on energy-independent Q-values, to a detailed method based on NJOY. A particular feature is in hybrid methods that consider energy dependence employed in NJOY in the system. The hybrid methods provide tools to assess errors in the simple methods and to evaluate the heat transport property by neutrons and photons precisely. The accuracy of the hybrid methods and the errors in the simple methods are investigated in a Monju design calculation. It is confirmed that the hybrid methods realize the nuclear heating by the detailed method within an accuracy of 0.1%. The neglect of incident energy dependence of the fission Q-value in the simple methods causes clear overestimation in the Q-value but its influence on the total heating is less than 1%.

Criticality evaluation for the Monju restart core

Hazama, Taira; Kitano, Akihiro; Kishimoto, Yasufumi*

Nuclear Technology, 179(2), p.250 - 265, 2012/08

https://doi.org/10.13182/NT12-A14097

The Japanese prototype fast breeder reactor Monju restarted its system startup test in May 2010 after a 14-year interruption. In the first stage of the test, reactor physics parameters have been measured at a zero power level. The present paper describes the evaluation of the criticality data. The best-estimate value and its uncertainty are evaluated as accurately as possible. The restart core contains 1.5 wt% of Am which is three times larger than the previous test. To extract an influence of the Am accumulation on calculation accuracy, criticality data obtained in the previous test is evaluated in the same level of detail. The calculation accuracy is investigated with four major nuclear data libraries. It is confirmed that the accuracy is within 0.3%, 2 value of experimental uncertainty, with JENDL-3.3, JENDL-4.0, and ENDF/B-VII.0. The reactivity change due to the Pu decay can be simulated within an accuracy of 1% with JENDL-4.0 and JEFF-3.1.