Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
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.
Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 10 Pages, 2023/10
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.
Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki
JAEA-Technology 2022-030, 80 Pages, 2023/02
JAEA-Technology-2022-030.pdf:2.57MBJAEA-Technology-2022-030(errata).pdf:0.11MB
Nuclear criticality benchmark analyses were carried out for TRIGA-type reactor systems in which uranium-zirconium hydride fuel rods are loaded by using the continuous-energy Monte Carlo code MVP with the evaluated nuclear data library JENDL-5. The analyses cover two sorts of benchmark data, the IEU-COMP-THERM-003 and IEU-COMP-THERM-013 in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook, and effective neutron multiplication factors, reactivity worths for control rods etc. were calculated by JENDL-5 in comparison with those by the previous version of JENDL. As the results, it was confirmed that the effective neutron multiplication factors obtained by JENDL-5 were 0.4 to 0.6% greater than those by JENDL-4.0, and that there were no significant differences in the calculated reactivity worths by between JENDL-5 and JENDL-4.0. Those results are considered to be helpful for the confirmation of calculation accuracy in the analyses on NSRR control rod worths, which are planned in the future.
Motome, Yuiko; Akiyama, Yoshiya; Murao, Hiroyuki
Journal of Nuclear Engineering and Radiation Science, 6(2), p.021115_1 - 021115_11, 2020/04
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.
Motome, Yuiko; Akiyama, Yoshiya; Murao, Hiroyuki
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
The NSRR is a research reactor of TRIGA-ACPR type, located in the Nuclear Science Research Institute. 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 after the Fukushima Daiichi accident, the research reactors are being regulated according to the risk of the facility. Graded approach is introduced in the regulation. In order to apply the graded approach, the radiation effects of residents living around the NSRI under the external hazards were evaluated and the level of the risk of the NSRR facility was investigating. This report is summarized for the result of the evaluation in case the safety functions were lost by the tornado, earthquake and following tsunami. As the result, the risk is confirmed to be low, since the effective dose of the residents has been below 5 mSv per event due to the loss of the safety functions by the tornado, earthquake and following tsunami.
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.
Suzuki, Motoe; Fuketa, Toyoshi; Saito, Hiroaki*
Nuclear Technology, 155(3), p.282 - 292, 2006/09
Times Cited Count：18 Percentile：75.30(Nuclear Science & Technology)Experimental analyses were performed for the RIA-simulated tests, OI-10 and OI-11 of high burnup PWR rods, in the NSRR by the RANNS code. The rod conditions were calculated by the fuel performance code FEMAXI-6 following the actual power history from the beginning to the end of irradiation in PWR and the results were given to the RANNS code as pre-test conditions. The RANNS analysis was conducted on the basis of such test conditions in the NSRR as the pre-test conditions, pulse power enthalpy and coolant temperature. The predicted quantities such as temperature of pellet stack and cladding, stress-strain distribution in cladding, and interactions among them during pulse irradiation were discussed in terms of PCMI process and compared with the experimental observations. In the OI-10 rod, calculated cladding permanent strain has a reasonable agreement with strain profile obtained in PIE, while locally enhanced strain of cladding was pointed out. In the OI-11 rod, the process from crack initiation to split failure was accounted for by the plastic strain occurrence in cladding.
Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi
Nuclear Engineering and Design, 236(2), p.128 - 139, 2006/01
Times Cited Count：8 Percentile：48.89(Nuclear Science & Technology)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.
Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.579 - 601, 2005/10
The RANNS code analyzes behaviors of a single fuel rod in reactivity-initiated accident (RIA) conditions. The code has two types of mechanical model; one-dimensional deformation model for each axial segment length of rod, and newly-developed two-dimensional local deformation model for one pellet length. Analyses were performed on the two RIA-simulated experiments in the NSRR, OI-10 and OI-11 with high burnup PWR rods, and results of cladding deformation were compared between calculations by the two models and PIE data. RANNS calculated the deformation profiles of claddings during the power transient of the experiments on the basis of the pre-pulse conditions of rods predicted by FEMAXI-6 code. In the calculations by the two-dimensional model, the plastic strain increase at the cladding ridges was compared with those in between the ridges and with the PIE data, and effect of stress variance induced by local non-uniformity of strain on the crack growth was discussed.
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 
81
m, 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.
Tomiyasu, Kunihiko; Sugiyama, Tomoyuki; Nakamura, Takehiko; Fuketa, Toyoshi
JAERI-Research 2005-022, 128 Pages, 2005/09
JAERI-Research-2005-022.pdf:16.11MB
In order to clarify the driving force of PCMI failure on high burnup fuels and the influence of hydrogen embrittlement on failure limit under RIA conditions, simulated-RIA experiments were performed on fresh fuel rods in the NSRR. The driving force was restricted only to thermal expansion of pellet by using fresh pellets, and fresh claddings were pre-hydrided to simulate hydrogen absorption of high burnup fuels. In seven experiments, test rods resulted in PCMI failure, which was observed on high burnup fuels, in terms of transient behavior and fracture configuration. It indicates that the driving force is sufficiently explained with thermal expansion of pellet and a contribution of fission gas is small. Many incipient cracks were generated in the outer surface of the cladding, and they stopped at the boundary between hydride rim and metallic layer. It suggests that a toughness of metallic region except hydride rim has particular importantance for failure limit. Fuel enthalpy at failure correlates with the thickness of hydride rim, and tends to decrease with thicker hydride rim.
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.
Fuketa, Toyoshi; Nagase, Fumihisa; Sasahara, Akihiro*
Nihon Genshiryoku Gakkai-Shi, 47(2), p.112 - 119, 2005/02
Behavior of LWR fuel during reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA) is described.
Nagase, Fumihisa; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 42(1), p.58 - 65, 2005/01
Times Cited Count：50 Percentile：93.73(Nuclear Science & Technology)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.
Sasajima, Hideo; Sugiyama, Tomoyuki; Nakamura, Takehiko*; Fuketa, Toyoshi
JAERI-Research 2004-022, 113 Pages, 2004/12
JAERI-Research-2004-022.pdf:47.48MB
Results from power burst tests, GK-1 and GK-2, conducted at the NSRR, are summarized. The tests were performed on a 14
14 PWR fuel rod irradiated to a burnup of 42 MWd/kgU in the Genkai unit #1 of Kyushu Electric Power Co., Inc. The instrumented test fuel rod in a double-container-type capsule was subjected to the pulse-irradiation with stagnant water cooling condition at 0.1 MPa and 293 K. Deposited energy and peak fuel enthalpy were 505 J/g and 389 J/g in the Test GK-1, and 490 J/g and 377 J/g in the Test GK-2, respectively. During the pulse-irradiations, DNB occurred and the cladding surface temperature reached 581 K and 569 K in the Tests GK-1 and -2, respectively. The maximum cladding hoop strain was 2.7% in the Test GK-1 and 1.2% in the Test GK-2. However, the test fuel rods did not fail. Estimated fission gas releases during the pulse-irradiations were 11.7% and 7.0% in the Tests GK-1 and -2, respectively.
Sugiyama, Tomoyuki; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 41(11), p.1083 - 1090, 2004/11
Times Cited Count：11 Percentile：58.02(Nuclear Science & Technology)The effect of cladding surface pre-oxidation on the rod coolability under reactivity initiated accidents was investigated. NSRR tests on irradiated fuel rods have shown higher rod coolability than that of fresh rods, which arose from suppressed DNB and early quench at the surface. To identify the dominant factor, possible factors such as pellet cracking and so on, were assessed. The most probable factor, the cladding pre-oxidation, was examined by pulse irradiation tests on fresh rods with three cladding surface conditions, no oxide layer, 1m and 10m-thick oxide layers. Temperature measurements showed increased thresholds for DNB and quench at the pre-oxidized surface, leading to a reduced film boiling duration. The shifts of the critical and minimum heat flux points could be caused by the surface wettability increase. In the present tests, the wettability change was probably dominated by the chemical potential change at the surface due to pre-oxidation. The test results indicate the effects do not depend on the oxide layer thickness, but on the presence of the oxide layer.
fuel at burnup of 45 GWd/t during simulated Reactivity Initiated Accident (RIA) conditionAmaya, Masaki; Sugiyama, Tomoyuki; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 41(10), p.966 - 972, 2004/10
Times Cited Count：8 Percentile：48.24(Nuclear Science & Technology)Pulse irradiation simulating RIA condition was carried out for test rod prepared from fuel irradiated in a commercial reactor. After the pulse irradiation, optical microscopy (OM) and scanning electron microscopy (SEM) observations and electron probe micro analysis (EPMA) were conducted for the test rod as a part of destructive tests. Fission gas release behavior during pulse irradiation was investigated by EPMA and puncture test. Xeon depression was observed in the fuel pellet after pulse irradiation at periphery and center region. It is considered that fission gas was mainly released from the pellet center region during pulse irradiation. The amount of xenon release during pulse irradiation was estimated to be 10-12% from the EPMA results and this estimated value was comparable with the puncture test result. Comparing the estimated value with other results of out-of-pile annealing tests, it was concluded that most fission gas, which was accumulated at grain boundary during base irradiation, was released from the center region of test fuel pellet during pulse irradiation.
Sugiyama, Tomoyuki; Fuketa, Toyoshi; Ozawa, Masaaki*; Nagase, Fumihisa
Proceedings of 2004 International Meeting on LWR Fuel Performance, p.544 - 550, 2004/09
Two pulse irradiation experiments simulating reactivity initiated accidents were performed on high burnup (60 GWd/t) PWR UO
rods with advanced cladding alloys. Test OI-10 was performed on an MDA cladded rod with large-grain (25 m) fuel pellets with a peak fuel enthalpy condition of 435 J/g, and resulted in a peak residual hoop strain of 0.7%. On the other hand, Test OI-11 on a ZIRLO cladded rod with conventional pellets resulted in a fuel failure at a fuel enthalpy of 500 J/g due to the pellet-cladding mechanical interaction (PCMI). A long axial split was generated on the cladding over the active length. The fuel pellets were fragmented and dispersed into the coolant water. The fuel enthalpy at failure is higher than the PCMI failure criterion of 209 J/g at the corresponding burnup. The experimental results suggest that the rods with improved corrosion resistance have much safety margin against the PCMI failure compared to the conventional Zircaloy-4 rod.
Sugiyama, Tomoyuki; Nagase, Fumihisa; Nakamura, Jinichi; Fuketa, Toyoshi
HPR-362, Vol.2, 12 Pages, 2004/05
To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as loss of coolant accident (LOCA) and reactivity-initiated accident (RIA) is being studied at the Japan Atomic Energy Research Institute (JAERI). The LOCA program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. The tests on irradiated claddings have recently been started and results have been obtained. As for an RIA study, a series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the LOCA and RIA studies at JAERI.
Murao, Hiroyuki; Yachi, Shigeyasu; Ota, Kazunori; Muramatsu, Yasuyuki; Nakamura, Takehiko; Terakado, Yoshibumi
UTNL-R-0435, p.15_1 - 15_9, 2004/03
no abstracts in English
