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; Sato, Wakaei*;
JNC TN9400 2000-037, 87 Pages, 2000/03
JNC-TN9400-2000-037.pdf:3.48MB
ln order to compare the nuclear characteristics of water-cooled bleeder cores with that of LMFBR, MOX fuel cell models are established for boiling and non-boiling LWR, non-boiling HWR and sodium-cooled reactor. Frst, the comarison is made between the heterogeneous cell calculation results by SRAC and those by SLAROM. The results show some differences as for neutron energy spectrum, one-grouped cross section and conversion ratio due to the different grouped cross section library (both are based on JENDL-3.2, though) used for each code, however, the difference is acceptably small for grasping the basic characteristics of the above-mentioned cores. Second, using the SLAROM code, main core parameters such as mean neutron energy, ratio of fast neutron and 
-value, are analyzed. The comparison between the cores show that softened neutron spectrum by the scattering effect of hydrogen or heavy hydrogen increase the contribution of nuclear reaction (especially for neutron capture reaction rather than fission reaction) in lower energy region comparing with LMFBR. ln order to overcome the effect, tighter lattice than LMFBR is necessary for water-cooled cores to realize the breeding of fissile nuclides. Third, effects of Pu isotopic composition on the breeding ratio are evaluated using SRAC burnup calculation. From the results, it is confirmed that degraded Pu (larger ratio of Pu-240) show the larger breeding ratio. At last, sensitivity analyses are made for k-effective and main reaction ratios. As for k-effective, using a temporary covariance data of JENDL-3.2, uncertainty resulting from the cross sections' error is analyzed for a boiling LWR and a sodium-cooled reactor. The boiling LWR core shows larger sensitivity in lower energy region than the sodium-cooled reactor (especially for the energy region lower than 1kev), And, 18-group analysis that is considered acceptably good for LMFBR analysis, should not be enough for accurate sensitivity estimation of ...
;
JNC TN9400 2000-034, 48 Pages, 2000/03
JNC-TN9400-2000-034.pdf:1.56MB
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.
*; *; *; *
JNC TJ9440 2000-007, 43 Pages, 2000/03
JNC-TJ9440-2000-007.pdf:1.73MB
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.
Koshizuka, Seiichi*
JNC TJ9400 2000-011, 102 Pages, 2000/03
JNC-TJ9400-2000-011.pdf:2.71MB
In order to evaluate the possibility to achieve high electric power by a fast reactor with supercritical light water, the design study was carried out on a large fast reactor core with high coolant outlet temperature (SCFR-H). Since the reactor coolant circuit uses once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure, it is possible to design much simpler and more compact reactor systems and to achieve higher thermal efficiency than those of current light water reactors. The once-through direct cycle system is employed in current fossil-fired power plants. In the present study, three types of core were designed. The first is SCFR-H with blankets cooled by ascending flow, the second is SCFR-H with blankets cooled by descending flow and the third is SCFR-H with high thermal power. Every core was designed to achieve the thermal efficiency over 43%, positive coolant density reactivity coefficient and electric power over 1600MW. Core characteristics of SCFR-Hs were compared with those of SCLWR-H (electric power: 1212MW), which is a thermal neutron spectrum reactor cooled and moderated by supercritical light water, with the same diameter of the reactor pressure vessel. It was shown that SCFR-H could increase the electric power about l.7 times maximally. From the standpoint of the increase of a reactor thermal power, a fast reactor has advantages as compared with a thermal neutron reactor, because it can increase the power density by adopting tight fuel lattices and eliminating the moderator region. Thus, it was concluded that a reactor cooled by supercritical light water could further improve the cost competitiveness by using a fast neutron spectrum and achieving a higher thermal power.
Otagaki, Takao*
JNC TJ1420 2000-004, 159 Pages, 2000/03
JNC-TJ1420-2000-004.pdf:4.82MB
no abstracts in English
*; Kitada, Takanori*; Tagawa, Akihiro; *; Takeda, Toshikazu*
JNC TJ9400 2000-006, 272 Pages, 2000/02
JNC-TJ9400-2000-006.pdf:9.69MB
Investigation was made on the follwing three themes as a part of the improvement of reactor physics analysis method for FBR with various core concept. Part 1: Investigation of Error Estimation of Neutron Spectra in FBR and Suggestions to Improve the Accuracy. In order to improve the spectrum unfolding method used in fast experimental reactor JOYO, a trial was made to evaluate the error in the estimated neutron spectrum, cause by cause. And the evaluated errors were summed up to obtain the most probable and reasonable error as possible. The summed up error was found relatively small compared to the error caused by the uncertainty of cross section data: most of the error in the spectrum unfolding method can be attributed to the error in cross sections. It was also found that the error due to the fission spectrum causes a considerable error in the high energy neutron spectrum which is over several MeV. Part 2: Study on Reactor Physics Analysis Method for Gas-Cooled FBR. In gas-cooled FBR, the portion of coolant channels in core volume is larger than sodium-cooled FBR. This leads to strong neutron streaming effects. For sodium-cooled FBR, several methods were proposed to evaluate the neutron streaming effect, however, these methods can not be used directly to gas-cooled reactor because the direction dependent diffusion coefficient becomes infinitive along the direction pararel to the coolant chammel. In this study, a new method is proposed to evaluate the neutron streaming effect, based on the method taking the axial buckling into consideration, which method was originally proposed by K
hler. Part 3: Study on Reactor Physics Analysis Method for Water-Cooled FBR An investigation was made on low-moderated water-cooled FBR, on the point that the ordinary used analysis method for FBR may give considerable difference in results in such core. In light water reactors, it is well known that the space dependence of self-shielding effect of heavy nuclides are considerably ...
*; *; Tanai, Kenji
JNC TN8400 99-047, 54 Pages, 1999/11
This paper reports on the design process for a carbon-steel overpack as a key component in the engineered barrier system of a deep geological repository described in the 2nd progress report. The results of the research and development regarding design requirements, configuration, manufacturing and inspection of overpack are also described. The concept of a composite overpack composed of two different materials is also considered. First, the design requirements for an overpack and presume environmental and design conditions for a repository are provided. For a candidate material of carbon steel overpack, forging material is selected considering enough experience of using this material in nuclear power boilers and other components. Second, loading conditions after emplacement in a repository are set and the pressure-resistant thickness of overpack is calculated. The corrosion thickness to achieve an assigned 1000 year life time and the required thickness to prevent radiolysis of ground water which might enhance corrosion rate are also determined. As aresult, the total required thickness of a carbon-steel overpack is conservatively estimated to 190 mm. This is a reduction of about 30% from the previous estimate provided in the 1st Progress Report. Additional items that must be considered in manufacturring and operating overpacks (i.e. sealing of vitrified waste, examination of main body and sealing welding, mechanism of handling) are evaluated on the basis of current technology, specific future data needs are identified. With respect to the concept of composite overpack (i.e., an outer vessel to provide corrosion-allowance or corrosion-resistant performance and an inner vessel to provide pressure-resistance), the differences in design concepts between the carbon-steel overpack and such composite overpacks are analyzed. Future data needs and analytical capabilities with respect to overpacks are also summarized.
Kofuji, Hirohide;
PNC TN9410 98-066, 67 Pages, 1998/06
The economics of various fuel cycle based on the current domestic situation were evaluated to confirm the adequacy of the economical target of the FBR fuel cycle. Four types of fuel cycle concepts, LWR-once through, LWR-reprocessing, LWR-Pu and FBR, were selected. The unit costs of the fuel cycle components were based on the domestic business base price or estimated values from published construction costs. These setting of the unit costs were the key point of this report. As the results, generating electricity cost of the LWR-reprocessing scenario which was thought to be the domestic reference scenario was estimated to be 8.2 yen/kWh, that of once through scenario was 7.5 yen/kWh, and LWR-Pu scenario was 8.2 yen/kWh. These differences of the generating electricity costs due to the deference of the fuel cycle costs because those three scenario were based on the same reactor concept. The generating electricity cost of the once through scenario is smaller than that of the reprocessing scenario equivalent for the reprocessing cost because that front end costs of the both scenario were equal and spent fuel or waste disposal cost occupied merely small parts of the fuel cycle costs. In the PWR-Pu scenario, natural uranium and enrichment cost were cut down but the fuel cycle cost was the same as the reprocessing scenario due to the relatively high cost of MOX fabrication. On the other hand, though the gcnerating electricity cost of the FBR scenario based on the current technology is 13.3 yen/kWh which is about 1.6 times of that of the LWR reprocessing scenario, its future cost was estimated to be 6.7 yen/kWh which is lower than every LWR scenario results of the future R&D. These economical evaluation showed the adequacy of the economical target of the FBR cycle, it is necessary to improve the evaluation accompanying with the trend of the LWR cycle and the progress of the R&D of the FBR cycle.
Mori, Tomoaki;
PNC TN9410 98-056, 72 Pages, 1998/06
The sub-criticality monitoring system has been developed for criticality safety control in nuclear fuel handling plants. In the past experiments performed with the Deuterium Critical Assembly(DCA), it was confirmed that the detection of sub-criticality was possible to k
=0.3. To investigate the applicability of the method to more generalized system, experiments were performed in the light-water-moderated system of the modified DCA core. From these experiments, it was confirmed that the prompt decay constant(
), which was a index of the sub-criticality, was detected between k=0.623 and k 0.870 and the difference of 0.05
0.1
k could be distinguished. The values were numerically calculated with 2D transport code TWODANT and monte carlo code KENO V.a, and the results were compared with the measured values. The differences between calculated and measured values were proved to be less than 13%, which was sufficient accuracy in the sub-criticality monitoring system. It was confirmed that Feynman- method was applicable to sub-critical measurement of the light-water-moderated system.
Kofuji, Hirohide;
PNC TN9410 97-100, 49 Pages, 1997/10
The degradation of Plutonium isotopic composition is suggested by the multi-recycling in the LWR. On the other hand, it is expected that FBR or FR has some advantages from the view point of the use of the degraded Plutonium. In this report, the Plutonium mass flow was calculated on some scenarios focused on the trend of the Plutonium isotopic composition through several times recycling. As the results, the Plutonium composition was remarkably degraded in the case of LWR only recycling, however it would be recovered by using both the FBR core and the blanket fuels. In the case of FR recycling, Plutonium can be consumed steadily by using one ratio of LWR, LWR(Pu) and FR. Though the FBR system has some merits, for example saving the natural Uranium resource, it became clear that the FBR can be used for the purpose of using degraded Plutonium.
; ; Yoshikawa, Shinji; Ozawa, Kenji
PNC TN9410 96-101, 40 Pages, 1996/04
Since it is desired to enhance availability and safety of nuclear power plants operation and maintenance by removing human factor, there are many researches and developments for intelligent operation or diagnosis using artificial intelligennce (AI) technique. We have been developing an autonomous operation and maintenance system for nuclear power plants by substituting AI's and intelligent robots. It is indispensable to use various and large scale knowledge relative to plant design, operation, and maintenance, that is, whole life cycle data of the plant for the autonomous nuclear power plant. These knowledge must be given to AI system or intelligent robots adequately and opportunely. Moreover, it is necessaly to insure real time operation using the large scale knowledge base for plant control and diagnosis performance. We have been studying on the large scale and real time knowledge base system for autonomous plant. In the report, we would like to present the basic concept and expecting performance of the knowledge base for autonomous plant, especialy, autonomous control and diagnosis system.
PNC TN1100 93-008, 26 Pages, 1992/12
FBR development activities in Japan have been performed by the government in cooperion with private enterprises. The prototype reactor "MONJU" is now undergoing functnal testing, and the first demonstration plant is in the conceptual design stage. R for commercial plants has been conducted for several years. Commercial plants are quired to be superior to LWRs with regard to economy, safety, and reliability. Accoingly, the Power Reactor & Nuclear Fuel Development Corporation and the Japan Atomicower Company set up specific R&D goals for commercialization, identified plant conces, and planned the necessary R&D activities. In order to make the concepts a realit both government and private enterprises must play a roll in developing and demonstring FBR technologies through construction and operation of prototype and demonstrati plants. In addition, they must perform FBR optimization activities, such as an enhced safety core and a passive decay heat removal system etc., according to lo