Bunseki, 2019(10), p.459 - 461, 2019/10
no abstracts in English
Chiba, Satoshi*; Wakabayashi, Toshio*; Tachi, Yoshiaki; Takaki, Naoyuki*; Terashima, Atsunori*; Okumura, Shin*; Yoshida, Tadashi*
Scientific Reports (Internet), 7(1), p.13961_1 - 13961_10, 2017/10
Transmutation of long-lived fission products (LLFPs: Se, Zr, Tc, Pd, I, and Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 10 to 10 years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contribute to developing a self-consuming cycle of LLFPs using fast spectrum reactors to reduce radioactive waste.
Kaji, Yoshiyuki; Matsui, Yoshinori; Kita, Satoshi; Ide, Hiroshi; Tsukada, Takashi; Tsuji, Hirokazu
Nuclear Engineering and Design, 217(3), p.283 - 288, 2002/09
In the Japan Atomic Energy Research Institute (JAERI), in-pile strain measurement techniques have been developed using the Japan Materials Testing Reactor (JMTR). In order to evaluate the performance of fiber optic grating sensors under irradiation environment, heat-up and performance tests at elevated temperatures before irradiation and in-pile tests were performed in JMTR. It was determined that it is possible to measure strain under irradiation environment below 11023n/m (E1MeV) by a fiber optic grating sensor, because in-pile temperature characteristics were in good agreement with out-of-pile test results.
AIP Conference Proceedings 561, p.223 - 231, 2000/09
no abstracts in English
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).
JNC-TN1440 2000-005, 214 Pages, 2000/08
no abstracts in English
; ; Sakamoto, Naoki; *; Akasaka, Naoaki;
JNC-TN9400 2000-095, 110 Pages, 2000/07
The effects of high fluence irradiation and swelling on the transient burst properties of austenitic steel fuel claddings; PNC316 and 15Cr-20Ni stcel, which were irradiated as the MONJU type fuel assemblies (MFA-1&MFA-2) in the FFTF reactor, were investigated. The temperature-transient-to-burst tests were conducted on a total of eight irradiation conditions. Fractographic examination and TEM observation were performed in order to evaluate the effect of high dose irradiation on the transient burst property and the relation between failure mechanism and microstructural change during rapid (ramp) heating. The results of the PIE showed that there was no significant effect of irradiation on the transient burst properties of these fuel claddings under the irradiation conditions examined. the results obtained in this study are as follows; (1)The rupture temperature of the irradiated PNC316 fuel cladding of MFA-1 was as same as that of our previous works for the fluence range up to 2.1310 n/m. There was no noticeable decrease in rupture temperature with increasing fluence in lower hoop stress region(100MPa). (2)The rupture temperature of the irradiated 15Cr-20Ni fuel cladding of MFA-2 was almost as same as that of as-received cladding for the hoop stress range up to about 200MPa. The rupture temperature did not decrease significantly with fluence. (3)The rupture temperature of the irradiated PNC316 cladding tested at hoop stress 69MPa, which was the design hoop stress for MONJU fuel, was 1055.6C. This suggested that the design cladding maximum temperature limit for MONJU (830C) was conservative. (4)There was no obvious relation between rupture temperature, swelling and microstructural change during transient heating under the irradiation conditions examined.
Matsubayashi, Masahito; Yoshii, Koji*; Hibiki, Takashi*; Mishima, Kaichiro*
Kashika Joho Gakkai-Shi, 20(Suppl.1), p.325 - 328, 2000/07
no abstracts in English
JNC-TJ9450 2000-002, 112 Pages, 2000/03
This report is intended to make it easier to apply the measured data obtained from the Gap Streaming Experiment, which was conducted at the Oak Ridge National Laboratory (ORNL) during about two months beginning at the start of March, 1992 as the sixth one of a series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) which was started in 1986. For this reason. the information presented includes specifications and measurement data for all configurations, compositions of all materials, characteristics of the measurement system. and daily-basis records of measurements. The Gap Streaming Experiment was planned to obtain the data of neutron streaming characteristics in the inclosure system above the core of an advanced fast reactor for verification and improvement of the analysis method to be applied to the shielding design. A iron-lined solid or slit concrete assembly was placed, with or without a spectrum modifier forming soft incident neutron spectrum, behind the TSR-II reactor of Tower Shielding Facility. Inserting central cylinders and cylindrical sleeves gave various gap width and offset in the slit concrete assembly. Neutron flux was measured behind the configurations with various types of detectors. The integral neutron flux in wide energy region was measured on radial traverse and on the axis behind the concrete assembly in almost all configurations. Neutron spectrum and fine radial distribution in high energy region was measured further in case of hard incident neutron spectrum, Information presented in this report is based mainly on a report issued by ORNL (ORNL/TM-12140. "Measurements for the JASPER Program Gap Streaming Experiment"). Additional information reported by the assignee is utilized also.
Mori, Tomoaki*; Takemura, Morio*
JNC-TJ9450 2000-001, 96 Pages, 2000/03
This report is intended to make it easier to apply the measured data obtained from the Special Materials Experiment, which was conducted at the Oak Ridge National Laboratory (ORNL) during about a month beginning at the end of June, 1992 as the last one of a series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) which was started in 1986. For this reason. the information presented includes specifications and measurement data for all configurations, compositions of all materials, characteristics of the measurement system. and daily-basis records of measurements. The Special Materials Experiment was planned to obtain the data of neutron attenuation characteristics of selected shielding materials for use in advanced fast reactors. The material of particular interest for the experiment was zirconium hydride that is rich in hydrogen. The mockup slabs for the special materials were preceded by the spectrum modifier behind the TSR-II reactor of Tower Shielding Facility. The layer of zirconium hydride was simulated with a combination of zirconium and polyethylene slabs. The thick layer of polyethylene with no zirconium was installed in some configurations.Neutron flux was measured behind the configurations with various types of detectors. The integral neutron flux in wide energy region was measured in eight configurations and neutron spectrum in high energy region was measured also in almost all configurations. Information presented in this report is based mainly on a report issued by ORNL (ORNL/TM-12277. "Measurements for the JASPER Program Special Materials Experiment"). Additional information reported by the assignee is utilized also.
Shiba, Tsuyoshi*; Kamezaki, Hiroshi*; Yuyama, Tomonori*; *
JNC-TJ9400 2000-012, 92 Pages, 2000/02
This research aims to develop a system in which aspects necessary for FBR cycle and overall comparison of evaluation items (economy, safety etc.) are evaluated quantitatively and objectively as a part of Nuclear Cycle development's research project of the FBR cycle for practical use. There are various methods in the decision-making support. In this particular situation, features of each method were evaluated based on the analysis of cases with each method. Subsequently we constructed overall evaluation method by combining Analytic Hierarchy Process (AHP), Multi-attribution Utility Function Method (MUF) and Cut-off Method. This method has variation in evaluation items, transparency in evaluation process and uncompensation. The six aspects of evaluation are economy, effectiveness of resource use, proliferation resistance, environmental effectiveness, safety, and research and development. The evaluation items and the evaluation index of each aspect were hierarchized and the evaluation structure was constructed. In the present study effect function for each evaluation index and pair comparison for examining significance of each item were utilized to select prospective systems for FBR cycle experimentally. The result confirmed reliability of our general assessment system as a decision-making support system for FBR system.
*; Kitada, Takanori*; Tagawa, Akihiro*; *; Takeda, Toshikazu*
JNC-TJ9400 2000-006, 272 Pages, 2000/02
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 Khler. 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 ...
Sato, Satoshi; Iida, Hiromasa; Plenteda, R.*; Valenza, D.*; Santoro, R. T.*
Fusion Engineering and Design, 47(4), p.425 - 435, 2000/01
no abstracts in English
PNC-TN9410 98-059, 53 Pages, 1998/06
The analysis program code STEDFAST; Space, TErrestrial and Deep sea FAST reactor ・gas turbine system; had been developed in PNC to get the best values of system parameters on fast reactor ・gas turbine power generation systems used as power sources for deep sea, space and terrestrial cogeneration. In this report, we performed a parameter survey analysis by using the code to study characteristics of the systems. Concerning the deep sea fast reactor ・gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor ・gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were perfomed on the base case of a Na cooled reactor of 40kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concening the terrestrial fast reactor ・gas tubine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100 C for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100MWt. In the comparison of calculational results for Pb and Na of primary coolant material, The primary coolant weight flow rate was naturally large for the fomer case compared with for the latter case because density is very different between them. ...
PNC-TN9410 98-044, 47 Pages, 1998/06
Thermal striping phenomena characterized by stationary random temperature fluctuations are observed in the region immediately above the core exit of liquid-metal-cooled fast breeder reactors (LMFBRs) due to the interactions of cold sodium flowing out of a control rod (C/R) assembly and hot sodium flowing out of adjacent fuel assemblies (F/As). Therefore the in-vessel components located in the core outlet region, such as upper core structure (UCS), flow guide tube, C/R upper guide tube, etc., must be protected against the stationary random thermal process which might induce high-cycle fatigue. In this study, thermal striping conditions at the tee junction in the MONJU EVST system (maximum temperature difference : 110 C, Velocity ratio between main and branch pipes : 0.25) were investigated numerically by the use of computer programs. From the investigations, the following results have been obtained: (1) Effects of the secondaly flows generated by the existence of 90 elbow located at upstream position of the tee junction were negligeble, because the flow velocity in the main pipe is 0.25 of the flow velocity in the branch pipe. (2) A ration between maximum and effective amplitudes of the temperature fluctuations calculated by the DINUS-3 code was 3.18. It was concluded that the value 6.0 as the ratio used in the integrity evaluation of the EVST system is a coservative side. (3) There was a limit in ability of a time-averaged multi-dimensional code AQUA, in the evaluation of thermal striping phenomena with recirculation flows. One of the reasons was considered that the local equilibrium of turbulence flows was not established in this tee junction problem.
; Tsukimori, Kazuyuki
PNC-TN9410 98-069, 128 Pages, 1998/05
There is a growing tendency to need structural analysis aided expert system, which adopts advanced analysis techniques and is useful adaptive design of large reactor. This report describes about development of the h-version adaptive mesh division function based on Yuge & Iwai method. From points of view about securing of analysis precision, reduction of work to make analysis data and decrease in calculation costs, to analyze smoothly the nonlinear problems is the main object of this system. The h-version adaptive mesh technique is the method that increases locally finite element mesh density, depending on dividing the elements that the absorbed energy quantity exceeds a standard value every increment step. We developed this h-version adaptive mesh division function and incorporate it in the general nonlinear finite element code. For the function this system has, we show the following. (1)It is possible to apply this system to the thermos elastic-plastic nonlinear problem. (2)The provided finite elements (a)4-Node Quadrilateral Plane-Stress Element (b)4-Node Quadrilateral P1ane-Strain Element (c)4-Node Quadrilateral Axisymmetric Solid Element (d)4-Node Layered Shell Element (3)The provided constitutive model (a)Ono-model (b)kinetic hardening rule (c)ORNL 10 cycles hardening rule (4)The repetition technique : Newton-Raphson technique (5)The application possible force type (a)The concentrated forces (b)The distributed forces (c)The self forces (d)The temperature forces (6)It is possible to apply the cyclic repetition force. (7)The dividing the elements technique (a)Rectifying the strain of the element shape depending on the aspect ratio (b)Dividing the elements that the absorbed energy quantity exceeds a standard value every increment step. (c)Add the function of input the plural absorbed energy quantity that is the estimate value of the division elements. The programming and giving the tests about this system was put into by RCCM.
; Moro, Satoshi; ; ; ;
PNC-TN9410 98-033, 284 Pages, 1998/03
System engineering division of OEC has being carried out a design study of the advanced nuclear fuel recycle system using electro-metallurgical process, aiming for improvements in safety, reliability, economy and a1so in environmental burden and nuclear non-proliferation. But the public criticism against nuclear power is more severe recently, and the situation is changing as seeing in the conclusion of the round-table conference on FBR. The researcher's meetings, in which researchers in PNC and from other organizations attended, were held during December, 1997 and March, 1998 in order to discuss on the advanced nuclear fuel recycle system and technology for FBR to be aimed in the future, and how to execute its research & development, etc. The conclusions of this meeting are as follows: (1)The future advanced FBR fuel cycle system shall be the system which has high potential for maximum utilization of uranium resources, and also for revolutionary improvements of economy, safety, environmental burden, etc. so as to be accepted in the society. (2)Regarding to the process of the future fuel eycle system, electro-metallurgical process that is able to apply for reprocessing of different types of fuel (oxide, metal and nitride) and is flexible for technical progress is recommended. Research & development of this system and technology shall be carried out. (3)The mission of PNC (new organization) is to select the most appropriate advanced FBR fuel cycle system from the viewpoint of the long-term FBR age in the future, and to conduct development of its system. It is expected for the new organization to execute its research and development steadily in cooperation with other research institutes, etc. under the nation-wide assessment and agreement. According to the above conclusions, the system engineering division will enhance the design study of the advanced FBR fuel cycle system and establish the definite concept of the system in cooperation with concerned in and ...
PNC-TJ9601 98-002, 115 Pages, 1998/03
In fuel cycles with recycled actinide, core characteristics are largely influenced by minor actinide (MA: Np, Am), Accurate nuclear data of MA such as fission cross section are required to estimate the effect of MA with high accuracy. In this study, fast neutron induced fission cross section of MA was measured using Dynamitron Accelerator in Tohoku University. New or improved techniques and tools with high precision and fast timing capability were developed for this study. Those are as follows: (1)Development of a sealed fission chamber,(2)Intensification of Li neutron target, (3)Improvement of time-resolution of Time-of-Right (TOF) electronic circuit, (4)Introduction of MA (Np237, Am241 and Am243) samples with large sample mass and (5)Introduction of a U235 sample with high purity. Using these improved tools and samples, fission cross section of Np237 was measured between 10 to 100 keV. On the other hand, averaged fission cross section for Maxwell distribution spectrum with kt=25.3 keV was measured for Am241 and Am243.
PNC-TJ9500 98-002, 126 Pages, 1998/03