Re-evaluation of nuclear criticality characteristics for infinite and finite heterogeneous lattice systems composed of uranium-zirconium hydride fuel rods used in the TRIGA annular core pulse reactor NSRR

Yanagisawa, Hiroshi; Motome, Yuiko

JAEA-Research 2025-010, 197 Pages, 2025/11

JAEA-Research-2025-010.pdf:3.5MB

For understandings of nuclear criticality risks of TRIGA fuel rods and review of safety measures for handling them, nuclear criticality characteristics for infinite and finite heterogeneous lattice systems composed of the NSRR fuel rods were re-evaluated with the use of a detailed computational model for the fuel rod. The MVP version 3 code was used with the JENDL libraries including the latest version, JENDL-5, for the re-evaluation. As the criticality characteristics, variations of neutron multiplication factors of the infinite and water-reflected finite systems were examined in detail with parameters of the lattice pitch and density of moderator water. From the results of the re-evaluated criticality characteristics, the minimum critical number of fuel rods for the water-reflected hexagonal shaped lattice system was obtained to be 46.8 0.2 using the JENDL-5 library. Moreover, the attainability of criticality without the water as moderator and reflector was examined because the zirconium hydride moderator and graphite reflector are equipped with the TRIGA fuel rod. It was found that the criticality is possible to be attained by 115.7 0.6 of the number of fuel rods, which is the smaller number of fuel rods than loaded in the NSRR standard core, even though no water exists.

Measurement of transient fission gas release from high-burnup MOX fuel under a simulated reactivity-initiated accident condition using fission gas dynamics testing technique

Taniguchi, Yoshinori; Urano, Kenta; Mihara, Takeshi; Udagawa, Yutaka; Kakiuchi, Kazuo; Katsuyama, Jinya

Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), p.1292 - 1301, 2025/10

Verification of the tables for control rods calibration at NSRR

Motome, Yuiko; Agake, Toshiki; Yanagisawa, Hiroshi

JAEA-Technology 2024-015, 30 Pages, 2025/01

JAEA-Technology-2024-015.pdf:1.36MB

The tables for calibration of control rods were verified, which is used positive period method and improved rod drop method of periodic inspection at Nuclear Safety Research Reactor (NSRR). Those tables are "DOUBLING TIME-REACTIVITY" and "DECAY OF NEUTRON FLUX AFTER INSTANTANEOUS REDUCTION OF REACTIVITY". They are prepared around 1975. Since those tables do not clearly express source of values and records of data used in calculations, the authors verified those tables again. For the verification, the tables were reproduced as follows. For the positive period method, the relationship between the period and reactivity was analytically evaluated by using the inhour equation with NSRR's parameters. For the improved rod drop method, the ratios of neutron flux after the rod drop with parameters of negative reactivities was calculated using the EUREKA- 2 code. As a result, the values described in the tables well agree with those by the present evaluation because it is confirmed that standard deviations of the differences in the value by between the present evaluation and the tables are less than 0.035%. For this reason, it is verified that these tables are valid in the practical use for NSRR operations.

Benchmark analyses on control rod worths of TRIGA reactor modeled in the ICSBEP handbook using continuous-energy Monte Carlo code MVP version 3

Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 10 Pages, 2023/10

https://doi.org/10.11484/jaea-conf-2024-001

The benchmark analyses on the TRIGA reactor of the IEU-COMP-THERM-013 (ICT-013) in the ICSBEP handbook using the MVP version 3 code were carried out with the Japanese, US and European nuclear data libraries, including JENDL-5. The analyses on effective neutron multiplication factors (k's) were also performed for another TRIGA reactor defined in the ICT-003 as well as the ICT-013. As a result, it is confirmed that the calculated k's vary in the range of 0.8% with the libraries. The results also suggest that there might be unknown bias in the ICT-013 for the calculated k's. For the analyses on the control rod worths of the ICT-013, it is confirmed that differences in the worths among the libraries become smaller than those in k's. Most of the errors involved in k's are considered to be cancelled in the calculation of the worths since the worths are obtained by subtraction of the reciprocals of two sorts of k's. The differences in the worths by between the benchmark and alternative methods defined in the ICT-013 were tried to be examined from the aspect of the differences in horizontal flux distributions by between those methods. It is confirmed that the differences in the worths for two shim control rods, which are fully withdrawn at a delayed critical state, are well understood by that attempt, but on the other hand the differences are not sufficiently explained for a regulating control rod, which is partially inserted to attain delayed criticality.

Evaluation of radiation effects on residents living around the NSRR under external hazards

Motome, Yuiko; Akiyama, Yoshiya; Murao, Hiroyuki

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021115_1 - 021115_11, 2020/04

https://doi.org/10.1115/1.4045075

The nuclear safety research reactor (NSRR) is a research reactor of training research isotopes general atomics -annular core pulse reactor type. The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity-initiated accident conditions. Under the new regulation standards, which was established after the Fukushima Daiichi accident, research reactors are regulated based on the risk of the facilities. To apply the graded approach, the radiation effects on residents living around the NSRR under the external hazards were evaluated, and the level of the risk of the NSRR facility was investigated. This paper summarizes the result of the evaluation in the case where the safety functions are lost due to a tornado, an earthquake followed by a tsunami. All in all, the risk is confirmed to be relatively low, since the effective dose on the residents is found to be below 5 mSv per event due to the loss of the safety functions.

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2013; Operation, utilization and technical development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

Department of Research Reactor and Tandem Accelerator

JAEA-Review 2014-047, 153 Pages, 2015/02

JAEA-Review-2014-047.pdf:23.43MB

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2013 and March 31, 2014.

Analysis on split failure of cladding of high burnup BWR rods in reactivity-initiated accident conditions by RANNS code

Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi

Nuclear Engineering and Design, 236(2), p.128 - 139, 2006/01

https://doi.org/10.1016/j.nucengdes.2005.07.009

A computer code RANNS was developed to analyze fuel rod behaviors in the RIA conditions. The code performs thermal and FEM-mechanical calculation for a single rod in axis-symmetric geometry to predict temperature profile, PCMI contact pressure, stress-strain distribution and their interactions. An experimental analysis by RANNS begins with pre-test conditions of irradiated rod which are given by FEMAXI-6. Analysis was performed on the simulated RIA experiments in NSRR, FK-10 and FK-12, of high burnup BWR rods in a cold start-up conditions, and PCMI process was discussed extensively. It was revealed that pellet thermal expansion dominates cladding deformation and subjects the cladding to bi-axial stress state, and thermal expansion in the cladding makes the stress in the inner region significantly lower than that in the outer region. Simulation calculations with wider pulses were carried out and the resulted cladding hoop stress was compared with the failure stress estimated in the NSRR experiments.

NSRR RIA-simulating experiments on high burnup LWR fuels

Fuketa, Toyoshi; Sugiyama, Tomoyuki; Sasajima, Hideo; Nagase, Fumihisa

Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.633 - 645, 2005/10

LWR fuel behaviors during a reactivity initiated accident (RIA) are being studied in the NSRR program. Results from recent NSRR experiments, no failures in Tests OI-10 and -12 and the higher failure enthalpy in Test OI-11, reflect the better performance of the new cladding materials in terms of corrosion during PWR operations. Accordingly, these rods with improved corrosion resistance have larger safety margin than conventional Zircaloy-4 rods. In addition, the smaller inventory of inter-granular gas in the large grain pellet could reduce the fission gas release in RIA as observed in the OI-10. Test VA-1 was conducted with an MDA sheathed 78 MWd/kgU PWR fuel rod. Despite of the higher burnup and thicker oxide layer of 81m, the enthalpy at failure remained in a same level as those for rods with of 40m-oxide at 50 - 60 MWd/kgU. This result suggests high burnup structure (rim structure) in pellet periphery does not have strong effect on the failure enthalpy reduction because the PCMI load is produced primarily by solid thermal expansion of the pellet.

RIA- and LOCA-simulating experiments on high burnup LWR fuels

Fuketa, Toyoshi; Nagase, Fumihisa; Sugiyama, Tomoyuki

Proceedings of IAEA Technical Meeting on Fuel Behaviour Modelling under Normal, Transient and Accident Conditions, and High Burnups (CD-ROM), 15 Pages, 2005/09

To provide a data base for the regulatory guide of light water reactors, behaviors of reactor fuels during off-normal and postulated accident conditions such as reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA) are being studied at the Japan Atomic Energy Research Institute (JAERI). A series of RIA-simulating experiments with high burnup fuel rods is being performed by using pulse-irradiation capability of the Nuclear Safety Research Reactor (NSRR). Fuel behaviors during a LOCA are also examined in an extensive program comprising of integral thermal shock tests and separate tests for oxidation rate and mechanical properties of fuel cladding.

Investigation of hydride rim effect on failure of Zircaloy-4 cladding with tube burst test

Nagase, Fumihisa; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 42(1), p.58 - 65, 2005/01

https://doi.org/10.3327/jnst.42.58

Tube burst tests have been performed with artificially hydrided Zircaloy-4 specimens at room temperature and at 620 K. Pressurization rate was increased to a maximum of 3.4 GPa/s in order to simulate rapid PCMI that occurs in high burnup fuel rods during a pulse-irradiation in the NSRR. Hydrogen content in the specimens ranged from 150 to 1050 ppm. Hydrides were accumulated in the cladding periphery and formed "hydride rim" as observed in high burnup PWR fuel claddings. The hydrided cladding tubes failed with an axial crack at the room temperature tests. Brittle fracture appeared in the hydride rim, and failure morphology was similar to that observed in the NSRR experiments. The hydrides rim obviously reduced burst pressure and residual hoop strain at the tests. The residual hoop strain was very small even at 620 K when thickness of the hydride rim exceeded 18% of cladding thickness. The present result accordingly indicates an important role of the hydrides layer in high burnup fuel rod failure under RIA conditions.