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Kinjo; Fukuhara; Oshita, Hironori; Takayama; Kitano, Akihiro; Takao; Yamazaki, Osamu*
JNC TN4410 2005-004, 243 Pages, 2005/09
; ; ; ; ; Kano, Shigeki
PNC TN9410 94-351, 97 Pages, 1994/09
Subsequent to the previuos testing (PNC SN9410 90-082, June 1990), characterization has been made on the sintered-metal load gauge element. The sintered-metal load gauge element was developed for use in off-line load measurement in the reactor environment. The testing conducted is as follows : (1)Characterization test phase II (a)Compression Tests for Initial Adjustment (b)Geometrical Parameter Compression Tests (c)Inclined Compression Tests (d)Creep Tests at Elevated Temperature (2)Characterization test phase III (For application in the reactor environment, the sintered-metal was covered with thin plates.) (a)Compression Tests for Initial Adjustment (b)Compression Tests at Elevated Temperatures (c)Inclined Compression Tests at Elevated Temperatures The results have shown that the sintered-metal load gauge element is applicable in the reactor environment. In association with the characterization tests, method for practical applications in JOYO and extended application have also been investigated.
Omura, Akiko; Yoshida, Akihiro; Shimakawa, Yoshio; Suzuki, Soju; Kinjo, Katsuya
PNC TN9410 93-290, 109 Pages, 1993/11
The "JOYO" modification program, named the MK-III program is now under investigation. The objective of this program is to make "JOYO" more efficient irradiation facility which serve as an irradiation bed for the development of FBR fuels and materials. During the 9th annual inspection, one of six control rods has been relocated from the 3rd row to the 5th row to improve the irradiation capacity, The control rod worth and several reactivity coefficients have been measured before and after the change of the control rod arrangement in detail. The principal results of the measurements are as follows; (1)In the rod drop measurements, the effect of the relative direction of the ex-vessel neutron instrumentation system and neutron source, which affects the accuracy of the measurements, was decreased by operating the measurements based on the experience that has obtained through the last Detailed Control Rod Worth Measurement. (2)The worth of the relocated control rod has decreased to 
1/3. (3)The other core neutronics characteristics were not changed by the change of the control rod arrangement. (4)The isothermal temperature reactivity coefficient for the core is -4
10
%
k/k/
C, which is almost same at every duty cycle. (5)It was confirmed that the flow rate reactivity coefficient consists of two independent factors, one may causes a reversible reactivity change and the other causes an irreversible reactivity change. Furthermore it is supposed that the threshold flow rate level that occurs an irreversible reactivity change is not the same throughout every duty cycle operation.
; Chatani, Keiji; ; ; ;
PNC TN9410 92-345, 166 Pages, 1992/10
An evaluation about the radioactive corrosion product (CP) behaviour in sodium cooling systems of a fast reactor is presented in this report, based on the obtained measurement results in the operating experience of JOYO. The objective of this work is to update the calculational model for predicting the release and deposition behaviour of CP in primary sodium cooling systems of a fast reactor. The evaluation results are as follows; (1)The main radionuclides of CPs transported to the out-of-reactor primary sodium loop are 
Mn and 
Co, and Mn is the most dominant. On the other hand, Co is the most dominant nuclide found in the liquid waste from spent fuel cleaning, which is produced by removal of activated CP deposits from surfaces of core sub-assemblies in sodium cleaning. (2)The deposition rate of Mn onto the hot-leg (HL) piping walls corresponds fairly with the saturation of radioactivity induced in core materials by activation, on the other hand, that onto the cold-leg (CL) piping walls has been being accelerated. The deposition rate of Co, due to the dependency of activation and release in a core, is strongly affected by the re-fuelling pattern and the oxygen concentration in sodium, and suggests the detouching process of deposits from wall surfaces. (3)Although Mn was transported and deposited preferentially in the HL of the primary cooling system in an early stage, the transport and deposition in the CL regions has overcomed that in the HL along operating time. Co was transported and deposited preferentially in the HL and the similar distribution pattern has been maintained thoroughout the operating periods. (4)The solution - precipitation model for CP behaviour in flowing sodium system was verified via the sensitivity test of model parameters and optimizing them on the above mentioned results, giving the measured to caluculated values of 1.36 or 1.03 for Mn or Co buildup, and 1.61 ...
; Chatani, Keiji; ; ;
PNC TN9410 92-224, 81 Pages, 1992/07
The radioactive corrosion product (CP) deposition density and gamma dose rate have been measured along the primary sodium circuits in Experimental Fast Reactor "JOYO" during every annual inspection and the CP behavior analysis code "PSYCHE" has been verified with measurement data in order to contribute the reduction of exposure dose of plant personal. The deposition density is measured by using a pure germanium detector system and determined by multiplying count rates by conversion factor. Gamma dose rate is measured with CaSO
thermoluminescence dosimeters (TLD). This report presents measurement results during the 9th annual inspection and the evaluation results for all data measured so far. The results on this study are summarized as follows: (1)Major CP nuclides deposited along the primary sodium circuits are Mn and Co. Mn is most dominant isotopes. Amounts of deposited Mn is about twenty times as much as those of Co. (2)Mn is deposited mainly on the cold leg pipings between the outlet of the intermediate heat exchanger (IHX) and the inlet of the reactor vessel. Co is deposited mainly on the hot leg pipings between the outlet of the reactor vessel and the inlet of IHX. (3)The buildup of Mn is saturated at 44.5 EFPY. The averaged dose rate of the pipings is saturated at about 1.5mSV/h. The dose rates of IHX and primary sodium pump are about 1.5 mSv/h and 2.1 mSv/h, respectively. The dose rate distributions around IHX and primary sodium pump show the peaks at the stagnant part of the flow and at the turbulence part. (4)Calculation by "PSYCHE" and measurement are compared. Calculation-to-measurement ratio is 1.2 for the CP deposition density and 1.5 for the dose rate. It can be said that the features of the CP behavior in the primary circuit of "JOYO" is made clear. The more effort will be required for the evaluation of CP behavior for subassemblies such as outer reflectors, clearness of ...
Chatani, Keiji; ; ; Masui, Tomohiko*; Nagai, Akinori; ;
PNC TN9410 92-186, 63 Pages, 1992/06
Dose rates around UGT (Upper Guide Tube) of CRDM (Control Rod Drive Mechanism) have been measured in Experimental Fast Reactor "JOYO" during the 9th periodical inspection in order to reflect the study on the shield thickness of UIS (Upper Internal Structure) cask, which has been planned to be used for a Large Fast Reactor. Absolute amount of radioactive corrosion products (CP) is evaluated by gamma spectra analysis for waste water from cleaned UGT. The results on this study are summarized as follows: (1)Measured dose rates distribution around UGT before and after clean-up show the same reduction. The affection of CP is not clearly observed for the dose rate distribution. (2)The relative values of dose rate, which are evaluated by considering the inside structure of UGT, show the attenuation of 10
from bottom to sodium level of UGT. The above relative distribution agrees well with that of measurement data using U-235 fission chamber, which was conducted at MK-I core start-up tests, except the stellite region. (3)As to the relative values of dose rate, calculation by "DOT3.5" and estimation by measured dose rate agree within factor 3 for the attenuation of 10. It is confirmed that the calculation can predict well the measurement. (4)Absolute amount of CP estimated by gamma spectra analysis and waste water analysis is 180 MBq. Co dominates 92 % of CP. This value agrees with the prediction by corrosion product behavior analysis code "PSYCHE" within factor 2.
; Chatani, Keiji; ;
PNC TN9450 92-001, 156 Pages, 1992/03
This report is intended to make it easier to apply the measured date obtained from the Axial Shield Experiment, which was conducted at the Oak Ridge National Laboratory (ORNL) in 1990 as part of a series of eight experiments planned for Japanese-American Shielding Program for Experimental Research (JASPER) program starting in 1986. The Axial Shield Experiment was planned to study the neutron attenuation characteristics of the axial shield, which is designed in the fuel assembly to reduce the neutron fluence in regions above the core. In order that the experimental neutron spectrum would be representative of the expected neutron spectra directly above the FBR core, the Tower Shielding Reacor (TSR) source spectrum was altered by a spectrum modifier, which was used in two previous experiments also. The modified spectrum entered the test section, which consisted of seven hexagonal shield assemblies surrounded by B,C and concrete. Three different axial shield designs were studies, Either B,C or stainless steel was used as a shielding material. Neutron measurements were made with various detectors behind the experimental configurations. Information presented in this report is based mainly on a report issued by ORNL (ORNL/TM-11839, "Measurements for JASPER Program Axial Shield Experiment"). Additional information reported by the PNC assignee is utilized also.
Chatani, Keiji; ; ; ; *; *; *
PNC TN9410 92-076, 348 Pages, 1992/03
JASPER (Japanese American Shielding Program of Experimental Researches) is the cooperative research program between PNC and US-DOE using TSF (Tower Shielding Facility) in ORNL (Oak Ridge National Laboratory) as the experiment facility. This report summarizes the works in FY'1991 as follows; Planning the experiment configuration for JASPER Program, Analyses of the JASPER Program experiment, Analyses of the former TSF experiment, and Development of the methods for FBR shielding analyses. (1)Analyses of the JASPER Program Experiment In FY'1991 Axial shield Experiment data were mainly analyzed, and some of In-vessel Fuel Storage (IVS) Experiment data were also analyzed. The Fast Reactor Shielding Analysis System developed by PNC has been applied to the analyses for JASPER Program experiments. (Axial Shield Experiment Analysis) Axial Shield Experiment was conducted from August 1990 through December 1990 as part of a continuing series of eight experimennts planned for the JASPER Program. The experiment serves not only to provide data for the verification of analysis system in calculating the neutron streaming in each design, but also to provide a basis for determining the shielding effectiveness of stainless steel (SS) and boron carbide (BC). four types of experimental configuration were used. The conclusions of the analyses are as follows: (a)For the spectrum modifier which provides a spectrum of neutron representative of those incident on the axial shield for the FBR core, the two-dimensional calculation showed good agreement with the experimental data. It was confirmed that the two-dimensional calculation could estimate the experimental data with almost the same accuracy as in the analyses for the Radial shield Attenuation and the Fission Gas Plenum Experiments. (b)For the homogeneous mockups, the two-dimensional ealculation could give the good agreement with the experimental data. (c)For the central blockage type mockups, in which the coolant flows ...
*; ; ; ; Fukami, Akihiro*; *; Igawa, Kenichi*
PNC TN9410 91-175, 52 Pages, 1991/05
An acoustic detection method is one of the FBR reactor core malfunction detection methods, and is regarded as being promising. In this study, the preliminary experiment of boiling detection by acoustic method was conducted at JOYO to measure the acoustic signal level and to investigate the applicability of the acoustic method. The experiment was performed on June 13 and 14, 1990 during the 8th periodic inspection of JOYO. The results obtained though the experiment are as follows: (1)Sodium bubbling (boiling) induced by the electric heater was detected as the fluctuation of temperature single of the thermocouple attached to surface of the electric heater. (2)Bubbling single of the acoustic detector could not be identified cleary because of the high background noise caused by the primary main pump vibration, sodium flow in the reacter vessel and the electric supply in the containment vessel. (3)The correlation between the signal of the acoustic detector or the fluctuation of temperature signal of the thermocouple and the flow rate of the primary loops was not ascertained. It became clear through this study that the validity of the reactor core malfunction detction by acoustic method depend on the peculiar noise level in the reactor vessel, and the reduction of noise is the subject for a future study.
; Arii, Yoshio; ; Suzuki, Soju;
3rd Asian Symposium on Research Reactor, 0 Pages, 1991/00
None
*; *
PNC TN9410 90-112, 76 Pages, 1990/07
This report su㎜arizes results on experiments, analyses and evaluations performed by Reactor Technology Section, Experimental Reactor Division during April through June, 1990. Each result described in this report was reported as the internal memoranda of Reactor Technology Section, for further analyses, evaluations, and/or discussions. This report contains the following items. (1)Results of measurements, analyses and evaluations for nuclear characteristics of JOYO. (2)Results of measurements and analyses for the plant characteristics of JOYO. (3)Production and/or arrangements of analyses codes and their manual. (4)Analyses and evaluations on MK- III Core. (5)Miscellaneous results. The final report will be published for each program after further discussions, analyses and evaluations.
*
PNC TN9410 89-171, 347 Pages, 1989/09
Based on the results of "Study on Main Design Parameters of Large Scale Fast Breeder Reactor" performed up to the fiscal year before last, an overall design study on a 1000 MWe-size FBR plant was made in the last fiscal year up to the level where cost evaluation could be done roughly. The design study was done on the assumption that the construction of the target plant be started within 1990's. The goal of this study was to design an FBR plant which was feasible to achieve a construction cost of 1.01.1 (formerly 1.11.2) times as much as that of light water reactor plant at the same age, and was also feasible to achieve an average burnup of 130,000MWd/t for discharged core fuel sub-assemblies. In this design study, however, the goals could not be achieved. It was made clear that further studies based on the results of the present study were necessary on main design parameters, and that more efforts were required for the fiscal year of 1989. As the result of this design study, problems were pointed out which should be taken into account in future R&D programs. And also, it was realized that the design study needs to be proceeded in cooperation with R&D activities.
Sanda, Toshio*; *; *
PNC TN9410 89-098, 278 Pages, 1989/06
Physics parameters were evaluated for FBR Cores with new fuels, i.e., metal, carbide and nitride fuels in comparison with the oxide fuel cores. The main results in FY'88 are as follows. (1)Some critical experiments and FBR core designs with new fuels were investigated and reviewed from the view point of nuclear characteristics of the core, For Zr and N cross sections, which are needed for new fuel cores, JENDL-3 library data must be used because these cross sections are not included in JENDL-2 library. (2)Physics parameters were compared between new fuels and oxide fuel cores. The conventional two-zone homogeneous core was selected for the calculation. The four cores were designed to be equal to another in dimension, core configuration, specific power density and peak linear heat rate, Comparisions are summarized as follows. (i)The burn-up reactivity loss is decreased to be nearly zero for the new fuel cores, and the breeding ratio is raiseed by about 0.2 for the new fuel cores compared with the oxide core. (ii)The maximum fast neutron fluence of the metal fuel core is about 30% larger than the oxide core due to harder spectrum and decreased effective fission cross section. (iii)The Na void reactivity and the Doppler coefficient for the metal fuel core are worse than those for the oxide core from the safety point of view, but reactivity coefficients due to fuel axial and radial expansion becomes more negative, (3)Cross section sensitivity analyses were conducted and uncertainties of core parameters due to cross section uncertainties were evaluated using cross section sensitivities and covariances. It was made clear that their total sensitivities to individual cross sections, except oxygan, were not very much different between the metal and oxide fuel core, though 100KeV1MeV sensitivities for the metal are larger than those for the oxide fuel core. The uncertainties of core parameters due to cross section uncertainties are almost equal for the ...
*; *; *; *
PNC TN9410 89-081, 136 Pages, 1989/04
We studied the concept of the small and medium-sized FBR as one of FBR development strategy. This year, we investigated the characteristics of reactor performances of the small and medium-sized FBR and the concept of innovative plant. Main results are as follows; (1)The characteristics of reactor performances of the small and medium-sized FBR (100 to 600MWe) was clear. (2)One problem of reactor performances of the small and medium-sized FBR is the reduction of the burn-up reactivity loss, because the smaller the reactor size is, the larger the burn-up reactivity loss is. (3)The reactivity coefficients of the small and medium-sized FBR, which is smaller size than 30kWe, is considerably different from large FBR. (4)According to quasi static approach, to have inherent shutdown capabilities, the linear heat rate should be reduced (below 250W/cm at 100MWe core) and the reactivity worth of one control rod should be limit to about 0.4%
k/kk'. (5)We established the concept of 100MWe innovative plant, making use of the characteristics of small reactor. The amount of core structure per power size of this plant is equal to Large FBR (1000MWe). Further, we should examine the amount of whole plant including BOP. Because of lower core pressure drop and linear heat rate, the capability of removing decay heat of this plant is superior to Large FBR. Because the fuel cycle cost of this plant is inferior to Large FBR (1000MWe), we should design high burn-up core.
*; *; *; Nakanishi, Seiji; *
PNC TN9410 89-041, 81 Pages, 1989/03
This work is on development of a method to evaluate the fuel maximum hot-spot temperature under the overpower situations at the beginning of burnup. The validity of the evaluation method is confirmed with the comparison of the hot-spot temperature obtained by this method and the probability density distribution of temperatures obtained by the Monte Carlo Method. The suitable pattern (pre-conditioning pattern) for increasing reactor power from zero to full power level is searched by using the method. It is also confirmed that the fuel melting probability is no greater than 0.03% by adopting the pattern. The features or the evaluation method are as follows: (1)The radial temperature increment from the cladding outer surface to the fuel center-line is evaluated by using the fuel performance analysis code CEDAR-III. (2)The reference temperature is defined to be the one evaluated under the conditions that the pellet density, equivalent fissile enrichment, cladding thickness and pellet densification are set to be certain upper levels and that the uncertainties of the reactor thermal power measurement and power distribution ealculation are taken into account. Here the growth of the central void is not overestimated by accounting the uncertainty of power distribution calculation only under the overpower situations excluding the normal operational situations. (3)The temperature uncertainties are evaluated due to the uncertainties of pellet outer diameter, Pu enrichment, O/M ratio, cladding inner diameter, relocation parameter and pore migration parameter which are independent of one another. (4)The total temperature uncertainty due to the set of the uncertainties are calculated by treating the individual temperature uncertainties statistically. The fuel maximum hot-spot temperature is given as the sum of the reference temperature and the total temperature uncertainty.
*; *; Nakanishi, Seiji; *
PNC TN9410 88-153, 113 Pages, 1988/12
This work is on the study of fuel cladding stress due to axial temperature difference at the boundaries of internal blanket(IB) sections, which is one of issues in fuel design field connected with the realization of axial heterogeneous core(AHC). Two-dimensional cladding temperature distribution near the boundaries of IB sections of AHC fuel pins is obtained by two-dimensional (axial and radial) thermal conduction analysis using Finite Element Nonlinear Structural Analysis System (FINAS). On the basis of this temperature distribution, cladding stress at the maximum linear heat rate level is evaluated by FINAS. As a result, it is found that cladding stress induced by thermal expansion rate difference at the beginning of irradiation gradually decreases with irradiation time owing to irradiation creep, but gradually increases inversely in and after the middle of irradiation when a swelling appears. The sign of cladding stress at the end of fuel life after 3 years irradiation duration is contrary to the one at the beginning, but the magnitudes of cladding stress are comparable to each other. Is also evaluated cladding stress due to radial temperature difference at the maximum linear heat rate level of homogeneous core(HOC), and the result is compared with the corresponding cladding stress in AHC. It is turned out that the difference of cladding stress between AHC and HOC is not so great. Consequently, it may be expected that fuel cladding stress due to axial temperature difference is not an obstacle to the realization of AHC. It should be mentioned that a cladding material is supposed to be an advanced austenitic steel(PNC1520, 15Cr-20Ni-2.5Mo-0.25Ti/0.1Nb).
