Aihara, Jun; Honda, Masaki*; Ueta, Shohei; Ogawa, Hiroaki; Ohira, Koichi*; Tachibana, Yukio
Nippon Genshiryoku Gakkai Wabun Rombunshi, 18(1), p.29 - 36, 2019/03
Japan Atomic Energy Agency carried out development of fabrication technology of oxidation resistant fuel element for improvement of safety of high temperature gas-cooled reactors in serious oxidation accident, based on precursor research in former JAEA. Dummy coated fuel particles (alumina particles) were over-coated with mixed powder of Si, C and small amount of resin to form over-coated particles, and over-coated particles were molded and hot-pressed to sinter dummy oxidation resistant fuel elements with SiC/C mixed matrix. We fabricated dummy oxidation resistant fuel elements with matrix whose Si/C mole ratio (about 0.551) is three times as large as that in precursor research. Si peak was not detected by X-ray diffraction of matrix. Better oxidation resistant was confirmed with oxidation test in 20% O at 1673 K than that of ordinal fuel compact with ordinal graphite/carbon matrix. All dummy coated fuel particles were held in specimen after 10 h oxidation.
; ; ;
JNC-TN8410 2001-021, 33 Pages, 2001/09
Carbon-14 released from the nuclear facilities is an important radionuclide for the safety assessment, because it tends to accumulate in environment through food chain and has as a significant impact to personal dose. Carbon-14 has been monitored routinely as one of the main gaseous radionuclides exhausted from the Tokai Reprocessing Plant (TRP) since OCtober of 1991. Furthermore, behavior of carbon-14 in TRP has been investigated through the reprocessing operation and the literature survey. This report describes the result of investigation about the behavior of carbon-14 in TRP as followings. (1)Only a very small amount of carbon-14 in the fuel was liberated into the shear off-gas and most of it was liberated into the dissolver of-gass. Part of the carbon-14 was trapped at the caustic scrubber installed in the of-gas treatment process, and untrapped carbon-14 was released into the environment from the main stack. Amount of carbon-14 released from the main stack was about 4.16.5GBq every ton of uranium reprocessed. (2)Carbon-14 trapped at the caustic scrubbers installed in the dissolver off-gas and in the vessel off-gas treatment process is transferred to the low active waste vessel. Amount of carbon-14 transferred to the low active waste vessel was about 5.4 9.6GBq every ton of uranium reprocessed. (3)The total amount of carbon-14 input to TRP was summed up to about 11.915.5 GBq every ton of uranium reprocessed considering the released amount from the main stack and the trapped amount in the off-gas treatment devices. The amount of nitrogen impurity in the initial fuel was calculated about 1522ppm of uranium metal based on the measured carbon-14. (4)The solution in the low active waste vesselis concentrated at the evaporator.Most of the carbon-14 in the solution was transferred into concentrated solution. (5)Tokai vitrification Demonstration Facility (TVF) started to operate in 1994. Since then, carbon-14 has been measured in the ...
JNC-TN1440 2000-007, 115 Pages, 2000/08
no abstracts in English
; Sato, Wakaei*; Iwai, Takehiko*
JNC-TN9400 2000-096, 113 Pages, 2000/06
This report describes the updated analyses results on the BFS-58-1-I1 core. The experiment was conducted at BFS-2 of Russian Institute of Physics & Power Engineering (IPPE). The central region is "non-Uranium fuel zone", where only Pu can induce fission reaction. The non-U zone is surrounded by MOx fuel zone, which is surrounded by U0 fuel zone. Sodium is used for simulating the coolant material. As it was found that the lattice pitch had been incorrectly understood in the past analyses, all items have been re-calculated using the corrected number densities. Furthermore, significantly softened neutron spectrum in the central region caused problems in applying the plate-stretch model that has been established for fast reactor cores through JUPITER experimental analyses. Both keeping the pellet density and using SRAC library for the elastic cross section for lighter nuclides allow us to obtain reasonable analysis accuracy on the spectral indices that were measured at the center of the core. Application of such a cell model was justified through comparison among various cell models using continuous energy Monte-Carlo code MVP. It is confirmed that both the MOX zone and the U0 zone can be correctly evaluated by the plate-stretch model. Based on the updated cell calculation, both the effective multiplication factor (k-eff)and the spectral indexes agree well with the measured values. The transport and mesh-size correction is made for the k-eff evaluation. Those results also agree well within reasonable difference between those obtained by IPPE and CEA, which were obtained by using sub-group method or continuous-energy Monte Carlo code. Evaluation by the nuclear data library adjustment confirmed that the analyses results of the BFS-58-1-I1 core have no significant inconsistency with JUPITER experimental analyses results. Those results are quite important for starting BFS-62 cores, which will be analyzed in the framework of supporting program for Russian ...
RANDEC Nyusu, (45), p.6 - 7, 2000/05
no abstracts in English
JNC-TN9400 2000-038, 98 Pages, 2000/04
As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.
; *; Nakazawa, Hiroaki;
JNC-TN8410 2000-012, 239 Pages, 2000/04
JNC has been conducted a great number of irradiation tests to develop MOX fuels for Advanced Thermal Reactor and Light Water Reactors. In order to manage irradiation data consistently and to effectively utilize valuable data obtained from the irradiation tests, we commenced construction of database system on MOX fuel for water reactors in 1998 JFY. Collection and selection of irradiation data and relevant fuel fabrication data, design of the database system and preparation of assisting programs have been finished and data registration onto the system is under way according to priority at present. The database system can be operated through the menu screen on PC. About 94,000 records of data on 11 fuel assemblies in total have been registered onto the database up to the present. By conducting registration of the remaining data and some modification of the system, if necessary, the database system is expected to complete in 2000 JFY. The completed database system is to be distributed to relevant sections in JNC by means of CD-R as a media. This report is an interim report covering 1998 and 1999 JFY, which gives the structure explanation and users manual concerning to the prepared database up to the present.
Oda, Yoshihiro; Funasaka, Hideyuki; *; Obara, Kenji*; *
JNC-TY8400 2000-002, 47 Pages, 2000/03
no abstracts in English
; Numata, Kazuyuki*
JNC-TN9400 2000-036, 138 Pages, 2000/03
Japan Nuclear Cycle Development lnstitute (JNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were renected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. ln this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor "JOYO" MK-l core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. 0n the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of "JOYO" MK-l core in comparison with ZPPR-9 core of JUPITER experiments.
JNC-TN9400 2000-034, 48 Pages, 2000/03
The study and the development to put FBR (Fast Breeder Reactor) to practical use have been doing. So many kinds of technologies are investigated to construct nuclear fuel recycle received to the society. The most important aim of reprocessing has been to extract U and Pu from spent fuels effectively, but, now, the demands for reprocessing are many kinds on nuclear fuel recycle system's construction. These need to be accepted sufficiently. The system that consists of electrolysis, extraction, with molten salt and melting metal, volatilization and condensation using the difference of vapor pressure is suggested, because, differently from LWR (Light Water Reactor), FBR can use the low decontamination factor's fuel. When the engineering scale plant is designed, the dry reprocessing has unsolved problems(ex. process flow) because of less demonstrative scale plants of the dry reprocessing than ones of the wet reprocessing. So the analysis and the estimation of mass balance that is most fundamental in the dry reprocessing system's design need to keep up with the system's alteration (to add new processes etc.) flexibly. This study aim is to develop the mass balance estimation code of dry reprocessing that satisfies the demand mentioned above.
*; Fujiwara, Masayuki*
JNC-TJ8430 2000-001, 55 Pages, 2000/03
no abstracts in English
*; *; *
JNC-TJ8420 2000-003, 99 Pages, 2000/03
no abstracts in English
JNC-TJ8400 2000-050, 47 Pages, 2000/03
In support of the safety assessment of geologic disposal of high levcl radioactive wastes, the solubility of transuranium elements was studied. The solubility of PuOxHO was measured undcr a reducing condition, and the solubility product K and the stability constant of Pu(OH) were obtained. The obtained K value was found to be much smaller than that predicted by Rai et al. from its dependence on ionic radius. Also, the solubility of PuO3 xHO was measured under an oxidizing condition and the solubility product K was obtained. In the analysis of hydrolysis constants of actinide ions, it was found that the systematic trend of the hydrolysis constants was well explained by the hard sphere model considering the effective charges of actinide ions.
JNC-TN9440 2000-002, 157 Pages, 2000/02
This report summarizes the operating and irradiation data of the experimental reactor "JOYO" 33rd cycle, and estimates the 34th cycle irradiation condition. Irradiation tests in the 33rd cycle are as follows: (1)B-type irradiation rig (B9) (a)High burn up performance tests of "MONJU" fuel pins, advanced austenitic steel cladding fuel pins, large diameter fuel pins, ferrite steel cladding fuel pins and large diameter annular pellet fuel pins (b)Mixed carbide and nitride fuel pins irradiation tests (in collaboration with JAERI) (2)C-type irradiation rig (C4F) (a)High burn up performance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (3)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (4)Absorber Materials Irradiation Rig (AMIR-6) (a)Run to absorber pin's cladding breach (5)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (6)Core Materials Irradiation Rig (CMIR-5-1) (a)Core materials irradiation tests (7)Structure Materials Irradiation Rigs(SMIR) (a)Material irradiation tests (in collaboration with universities) (b)Surveillance back up tests for "MONJU" (8)Upper core structure Irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect. The maximum burnup driver assembly "PFD516" reached 64,300MWd/t (pin average).
WILLIAM S.DERSHO*; *
JNC-TJ8440 2000-001, 408 Pages, 2000/02
J. A. BERRY*; M. BROWNSWORD*; D. J. ILETT*; Linklater, C. M.*; Mason, C.*; TWEED, C. J.*
JNC-TJ8400 2000-060, 60 Pages, 2000/02
Batch sorption experiments have been carried out to investigate the sorption behaviour of plutonium onto basalt and sandstone from the appropriate rock-equilibrated waters under different redox eonditions. Redox Potentials in solution were controlled by the addition of two reducing agents and one oxidising agent. Thermodynamic chemical modelling was undertaken to interpret the results. The sorption models were based on iron oxide. They adequately reproduced the data for sorption of plutonium onto sandstone, but tended to underpredict sorption onto basalt.
JNC-TJ8400 2000-044, 53 Pages, 2000/02
To estimate the polyelectrolyte effect and the effect of the heterogeneous composition of humic acids, the complex formation constants of Eu(III) and Ca(II) with Aldrich humic acid and polyacrylic acid were obtained, for Eu(10 to 10 M) by solvent extraction with TTA and TBP in xylene, for Ca (10M) with TTA and TOPO in cyclohexane and for Ca(10M) by using ion-selective electrode. By defining the apparent formation as = [MR]/([M][R]), where [R] denotes the concentration of dissociated functional group, [M] and [MR] denote the concentration of free and bound metal ion and pcH is defined as-log[H], the values of log have been obtained at pcH 4.8 - 5.5 in 0.1 - 1.0M NaClO and NaCl. Log of Eu-humate varied from 5.0 to 9.3 and that of Ca-humate from 2.0 to 3.4..For both humate and polyacrylate, log increased with pcH or with the degree of dissociation. The increase in the ionic strength O.1 to 1.0 M decreased the log, the decrease in log of Eu(III)-humate is 1.6, that of Eu(III), polyacrylate 0.7, that of Ca(II)-humate 1.9 and that of Ca(II)-polyacrylate 1.2. While the increase in the metal ion produced no effect on log of polyacrylate, log of humate decreased. Depending on the concentration of Eu(III), the coexistence of Ca(II) reduced log of humate by 0 to 0.8. The dependence of log of humate on the metal ion concentration suggests the coexistence of strong and weak binding sites in the hmnic acid.
PNC-TN1000 98-004, 21 Pages, 1998/07
no abstracts in English
PNC-TN9410 98-058, 12 Pages, 1998/06
Based on the RB1 test result in the CABRI-RAFT Program, it was agreed between the partners to perform the RB2 test which aims at observation of molten fuel ejection into the coolant channel at further fuel melting and at confirmation of coolability of ejected fuel. In this study, a preliminary post-test calculation for the RB1 test is performed first to reflect the fuel thermal condition expected for the pins with the special artificial defect preparation. Pre-test calculations for the RB2 test are then performed based on the results of this RB1 calculation. Power and coolant flow histories as well as the axial location of defect were selected as parameters in this study and a set of test condition is proposed which is believed to be most suitable to fulfill the test objectives.
PNC-TN9410 98-057, 55 Pages, 1998/05
Existing data of in-pile ramp-type transient-overpower tests (slow TOPs hereafter), such as those of the CABRI-2 and CABRI-FAST tests, the EBR-II TOPI-1E test and the former TREAT tests, were extensively surveyed and this led to a global interpretation which provided a consistency among the tests. Through this study, a basic fuel pin failure mechanism was comprehended and it was confirmed that fuel pins with low to intermediate smear density have a very high failure threshold with significant mitigation effects against fuel-cladding mechanical interactions. Such high failure threshold of low to intermediate smear density fuel is considered to be attributed to the following three effects: (1)absorption of fuel thermal expansion and fuel swelling by void space (porosity or cracks) within the fuel, (2)mitigation of fuel swelling by an early gas escape into the free volume, and (3)mitigation of molten cavity pressurization upon fuel melting. These effects were refrected to the analytical model of the transient fuel behavior code PAPAS-2S. Application of this improved PAPAS-2S model to representative slow TOP tests provided results consistent with the test data, suggesting that the above-mentioned consideration is valid.