Ohno, Shuji; Matsuki, Takuo*
JNC-TN9400 2000-106, 132 Pages, 2000/12
Sodium fire analyses were performed on 7 kinds of sodium leak tests using a computer code ASSCOPS which has been developed to evaluate thermal consequences of sodium leak accident in an FBR plant. By the comparison between the calculated and the test results of gas pressure, gas temperature, sodium catch pan temperature, wall temperature, and of oxygen concentration, it was confirmed that the ASSCOPS code and the parameters used in the analysis give valid or conservative results on thermal consequences of sodium leak and fire.
JNC-TN9440 2000-008, 79 Pages, 2000/08
This report summarizes the operating and irradiatlon data of the experimental reactor "JOYO" 35th cycle. Irradiation tests in the 35th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup performance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (2)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (3)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (4)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large scale reactor (5)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (6)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confimation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (based on a contract with universities) The maximum burnup driver assembly "PFD253" reached 67,600 MWd/t (pin average).
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JNC-TN8410 2000-011, 185 Pages, 2000/05
This report describes the neutronic design calculational methods used in Fuel Design and Evaluation Group in order to inform other related sections of FBR core analysis technology and hand down the technology. Especially we show the neutronics calculation procedures used for the conceptual design study of the advanced core with 127 pin bundle for MONJU that has been carried out in our group. The topics include effective cross section preparation calculations, two-dimensional depletion calculations, three-dimensional diffusion calculations, reactivity coefficient calculations, and control rod worth calculations. The calculational methods shown in this report are the standard neutronics calculation methods employed in our group at the moment. However, the improvement of calculation codes, the reduction of correction factors and uncertainties for design using the nuclear data obtained in the start-up test for MONJU and so on, and the update of nuclear data file will be planned in order to improve evaluation accuracies. Those may change the neutronic design calculational methods, but we decided to describe the present standard calculational methods in our group from the viewpoint of sharing information in JNC.
Ohtaki, Akira; ; ; *; *;
JNC-TN9410 2000-006, 74 Pages, 2000/04
To evaluate materials balance in nuclear fuel cycle quickly and quantitatively the fuel cycle requirement code "FAMILY" was improved. And an accumulated TRU&LLFP quantity analysis code was developed. The contents are as follows: (1)A calculation ability of minor actinide production and expenditure was added to the "FAMILY" code. (2)An output program for the "FAMILY" calculation results was developed. (3)A simple version of "FAMILY" code was developed. (4)An analysis code for accumulated TRU&LLFP quantity in nuclear fuel cycle was developed.
Fujiwara, Masayuki; Mizuta, Shunji;
JNC-TN9400 2000-050, 19 Pages, 2000/04
For evaluating the fast reactor system technology, it is important to evaluate the practical feasibility of ODS ferritic cdaddings, which is the most promising matelials to attain the goal of high coolant temperature and more than 150 GWd/t. Based on the results of their technology development, mass production process with highly economically benefit as well as manufacturing cost estimation of ODS ferritic claddings were preliminarily conducted. From the view point of future utility scale, the cost for manufacturig mother tubes has a dominant factor in the total manufacturing cost. The method to reduce the cost of mother tube manufacturing was also preliminarily investigated.
Mizuta, Shunji; ;
JNC-TN9400 2000-048, 28 Pages, 2000/04
ODS (Oxide Dispersion Strengthened) ferritic-martainsitic steels are one of the most prospective cladding materials for advanced fast breeder reactors, since they are expected to have excellent swelling resistance and superior high temperature strength due to the finely distributed stable oxide particles(YO). Properties and the tentative strength equations for ODS ferritic-martainsitic were proposed on the basis of the latest data to apply to the feasibility study of the sodium coolant MOX fuel plant. The items of equations are follows. (1)creep rupture strength (2)correction factor of creep rupture strength (in Na and in reactor) (3)outer surface eorrosion (Na) (4)inner surface corrosion (in MOX fuel pin) (5)thermal conductivity
Yamanaka, Shinsuke*; Uno, Masayoshi*; Kurosaki, Ken*; ; Namekawa, Takashi
JNC-TY9400 2000-011, 41 Pages, 2000/03
no abstracts in English
JNC-TN9400 2000-041, 29 Pages, 2000/03
Irradiation behavior and performance models were investigated in order to apply for nitride fuel options in feasibility study on fast breeder reactor and related recycle systems. (1)MechanicaI design of nitride fuel pin: The behaviors of fission gas release (increase of internal Pressure) and fuel-to-cladding chemical interaction (decrease of cladding thickness) are needed to evaluate cumulative damage fraction in case of fuel pin mechanical design. The behaviors of fission gas release and fuel-to-cladding chemical interaction were investigated from the past studies up to high burnuP, since the lower fission gas release in nitride fuel than in oxide fuel could contribute to reduce the plenum volume and result in the shortening of fuel Pin length. (2)Fuel pin smear density: The higher fuel smear density is preferred for the higher fissile density to improve the core characteristic. The behaviors of fuel pellet swelling were investigated from the past studies up to higher burnup, since the larger fuel pellet swelling in nitride fuel than in oxide fuel would restrict high burunp capability due to fuel-cladding mechanical interaction. (3)Compatibility of nitride fuel with high Temperature water: Compatibility of nitride fuel with high temperature water were investigated from the past studies to contribute water cooled fast breeder reactor options.
; Koyama, Tomozo; Funasaka, Hideyuki
JNC-TN8400 2000-016, 188 Pages, 2000/03
We summarized the conditions and results of all dissolution experiments (bench scale experiments (dissolution of sheared fuel pins) and beaker scale experiments (dissolution of a few sheared fuels pieces) of the irradiated fast reactor fuels, which were carried out in the Chemical Processing Facility (CPF). The fabrication and irradiation conditions of the dissolved fuels were also put in order.
JNC-TN8400 2000-015, 37 Pages, 2000/03
This report describes the study done within the period of time when I was postdoctoral research worker at Japan Nuclear Cycle Development Institute. The report includes two parts as follows. (1) Exact Solution of Electric Transitions for High Energy photons. Technologies for creating high-energy beams have been rapidly developed. These advancements make the research using high-energy -rays more important. The electric transition rates for high-energy -rays were formulated. The electric multipole fields were treated strictly in the process of calculating the electric transition rates and the nuclear states were taken as the harmonic oscillator wave functions. (2) Production of the isomeric state of Cs in the thermal neutron capture reaction Cs(n, )Cs. In order to obtain precise data of the neutron capture cross section of the reaction Cs(n, )Cs, the production probability of isomer state Cs was measured in this work. The 1436 keV -ray emitted from both of Cs and Cs was measured. A production ratio of Cs to (Cs and Cs) was deduced from time dependence of peak counts of 1436keV -ray. The probability of the production of CS was obtained as 0.750.18 and this value revised the effective cross section upwards 92%. The effective cross section and the thermal neutron capture cross section were obtained as =0.290.02 b and =0.270.03 b with taking into account the production of Cs.
; Koyama, Tomozo; Funasaka, Hideyuki
JNC-TN8400 2000-014, 78 Pages, 2000/03
We investigated the factors which affected the dissolution of U and Pu to the nitric acid solution with the fragmentation model, which was based on the results of dissolution experiments for the irradiated fast reactor fuels in the Chemical Processing Facility(CPF). The equation that gave the fuel dissolution rate was estimated with the condition of fabrication (Pu ratio (Pu/(U+Pu))), irradiation (burn-up) and dissolution (nitric acid concentration, solution temperature and U+Pu concentration) by evaluating these effects quantitatively. We also investigated the effects of fuel volume ratio to the solution in the dissolver, burn-up and flouring ratio of the fuel on the f-value (the parameter which shows the diffusion and osmosis of nitric acid to the fuel) in the fragmentation model. It was confirmed that the fuel dissolution rate calculated with this equation had better agreement with the results of dissolution experiments for the irradiated fast reactor fuels in the CPF than that estimated with the surface area model. In addition, the efficiency of this equation was recognized for the dissolution of unirradiated U pellet and high Pu enriched MOX fuel. It was shown that the dissolution rate of the fuel slowed down at the condition of the high U-Pu concentration dissolution by the calculation of the dissolution behavior with this equation. The dissolution of the fuel can be improved by increasing the nitric acid concentration and temperature, but from the viewpoint of lowering the corrosion of the dissolver materials, it is desirable that the f-value is increased by optimizing the condition of shearing and stirring for the improvement of dissolution.
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JNC-TJ9440 2000-007, 43 Pages, 2000/03
Planning of the plutonium utihzation in the Light water thermal reactor has been investigated to evaluate scenario for FBR development. Plans for MOX fuel utilization in the ABWR including Ooma plant are studied, and information of high burnup fuels for a future BWR is summarized based on public documents. Nuclear compositions of the present burnup fuel (45,000MWd/t) and a high burnup fue (60,000MWd/t) have been evaluated using an open code: SRAC. Results of the study are follows; (1)Surveying the status of MOX fuel utilization. The status of MOX and UO fuel utilization in the present BWR and future BWR have been summarized based on public documents. (2)Evaluation of spent MOX and UO fuel composition. Nuclear compositions of spent MOX and UO fuels at 45,000MWd/t and 60,000MWd/t burnup have been evaluated and summarized for recycle scenarios by FBR.
PNC-TN1000 98-004, 21 Pages, 1998/07
no abstracts in English
PNC-TN9410 98-017, 21 Pages, 1998/02
"Injector Test (1) of PNC High Power Electron Linac" was mentioned in last year report. 100mA beam with pulse length 20s repetition rate 1Hz and 50mA beam with pulse 1ms 0.5 Hz had been accelerated to 3.0 Mev successfully. The chopper and prebuncher system had not been used in that test. Now this report put emphases on the chopper and prebuncher systems tests. A good energy spectrum had been achieved using the chopper and prebuncher systems. And 100mA beam current with pulse length 3ms and repetition rate 0.1Hz was accelerated to 3.0 MeV.
PNC-TN9460 98-001, 156 Pages, 1998/01
This document provides a description of a calculational route, used in the Reactor Physics Research Section for sensitivity studies and initial design optimization calculations for fast reactor cores. The main purpose in producing this document was to provide a description of and user guides to the calculational methods, in English, as an aid to any future user of the calculational route who is (like the author) handicapped by a lack of literacy in Japanese. The document also provides for all users a compilation of information on the various parts of the calculational route, all in a single reference. In using the calculational route (to model Pu burning reactors) the author identified a number of areas where an improvement in the modelling of the standard calculational route was warranted - the document includes a description of these changes. The calculational route makes use of several different computer programs. SLAROM calculates nuclear data from compositions, using either homogeneous or heterogeneous models. CITATION and MOSES do reactor burn-up and/or flux diffusion calculations; CITATION is used for 2D (RZ) calculations, whilst MOSES models 3D (hex-Z) geometry. PENCIL and CITDENS are essentially specialized versions of CITATION (PENCIL includes data preparation and other functions). MASSN calculates fuel cycle mass balances. PERKY performs perturbation and associated calculations, both 1'st order and exact perturbations. JOINT and RZOUT3 provide various dataset interface functions, including energy group condensation. Briefer descriptions of the calculational route are given, followed by a more detailed step-by-step approach to the calculations. This latter includes examples of all JCL and data files, and a description of all the data that a user may have to employ. The document does not give a complete description of the component programs: where options and/or data are not used in any of the calculations they have generally been ignored; ...
PNC-TN1410 97-043, 167 Pages, 1997/11
no abstracts in English
; Hirosawa, Takashi
PNC-TN9410 97-075, 20 Pages, 1997/08
Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960's and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960's and that some of the 1960's data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO - PuO - PuO ideal solution model, and then formulized. The correlation obtained in this work is as-follows; T = T + T + T + T ----(A) T = 3120 T = -5.7537PU + 1.363110 PU + 1.795210 PU T = -1.41 PU (2.00 - OP)/0.39 OP : OP = /(0.01PU) T = -5.0BU/10000 where T is the melting temperature (degree of K), PU is the weight fraction of PuO in the mixed oxide fuel, OM is the oxygen to metal ratio, and BU is the burnup in the unit of MWd/MTM. respectively. T (plutonium content), T (O/M Ratio), ...
Onuki, Norihiko; ; Shuji, Yoshiyuki; ; ; ;
PNC-TN8410 97-272, 134 Pages, 1997/08
PNC-TN9410 97-064, 42 Pages, 1997/06
In our study on a hyblid fast reactor system, we have investigated it from the view point of transmutation ability of trans-uranium (TRU) nuclide making the most effective use of special features (controllability, hard neutron spectrum) of the system. It is proved that a proton beam is superior in generation of neutrons compared with an electron beam. Therefore a proton accelerator using spallation reaction with a target nucleous has an advantage to transmutation of TRU than an electron one. A fast reactor is expected to primarily have a merit that the reactor can be operated for a long temrm without employment of highly enriched plutonium fuel by using external neutron source such as the proton accelerator. Namely, the system has a desirable characteristic of being possible to self-sustained fissile plutonium. Consequently in the present report, core characteristics of the system were roughly studied by analyses using 2D-BURN code. The possibility of self-sustained fuel was investigated from the burnup and neutronic calculation in a cylindrical core with 300w/cc of power density without considering a target material region for the accelerator. For a reference core of which the height and the radius are both 100cm, there is a fair prospect that a long term reactor operation is possible with subsequent refueling of natural uranium, if the medium enriched (around 10wt%) uranium or plutonium fuels are fully loaded in the initial core. More precise analyses will be planed in a later fiscal year.
PNC-TN9410 97-057, 106 Pages, 1997/05
This study was based on a 'pancake' type fast reactor core design of 600 MW(e), which had been optimized for Pu burning with a feed Pu vector appropriate to once-through irradiation of MOX fuel in a PWR. The purpose of the study was to investigate the effects of varying the Pu vector, examining various methods of offsetting the effects of such a change, and finally to produce fuel cycles optimized for the different qualities of Pu vector within the same basic design. In addition to the reference (once-through) Pu vector, two extreme Pu vectors were examined: high quality Pu from military stockpiles; low quality Pu corresponding to the equilibrium point of multiple recycling in a Pu burning fast reactor. Variations in Pu quality were overcome by changing the fuel inventory - replacing some of the fuel by diluent material, and altering the fuel pin size. Using absorber material (BC) as diluent improves the rod worth shutdown margin but degrades the Na void and Doppler safety parameters, a non-absorber diluent has the opposite effects, so a mix of the 2 material types was used to optimize the core characteristics. Of the non-absorber diluent materials examined, ZrH gave significantly better performance than all others; BC was the second choice for non-absorber diluent, because of its compatibility with BC absorber. It was not possible to accommodate the lower quality (multi-recycled) Pu vector without a significant increase in the fuel pin volume. It was not generally possible, especially with the increased fuel pin size, to achieve positive rod worth shutdown margins - this was overcome by increasing the number of control rods. For the higher quality Pu vectors to maintain ratings within limits, it was necessary to adopt hollow fuel pellets, or else to use the diluent material as an inert matrix in the fuel pellets. It proved possible to accommodate both extremes of Pu vector within a single basic design, maintaining ...