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Okawachi, Yasushi; Maeda, Shigetaka; Nagasaki, Hideaki*; Sekine, Takashi
JNC TN9400 2003-029, 96 Pages, 2003/04
The (JOYO) MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition in 2001. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. However, after the database was published, it was recently found that there were errors in the process of making the group constant set JFS-3-J3.2, and lt was revised at JFS-3-J3.2R. Then, the group constant set was updated at JFS-3-J3.2R in this database. The MK-II core management data and core characteristics data were recorded on CD-ROM for user convenience. The structure of the database is the same as in the first edition. The (configuration Data) include the core arrangement and refueling record for each operational cycle. The (Subassembly Library Data) include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The (Output Data) contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The (Core Characteristics Data) include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. The effect of updating the group constant set, the calculation results excess reactivity decreased by about 0.15 delta-k/kk' , and the effects to other core characteristics were negligible.
Okawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi; Nagasaki, Hideaki*
JNC TN9400 2002-070, 49 Pages, 2003/01
JNC-TN9400-2002-070.pdf:1.78MB
As part of the JOYO upgrading program (MK-III program), the JOYO MK-III core management code system "HESTIA" was developed in order to improve the calculation accuracy concerning the core and fuel management and irradiation condition evaluation in the MK-III core. The neutronic calculation of HESTIA was modified to improve the power and neutron flux distribution. The calculation geometry was changed to Tri-Z geometry from Hex-Z geometry which was used in the MK-II core management code system "MAGI". The number of calculation mesh per subassembly was increased to 24 meshes in the radial direction, and the fuel region was divided into 20 meshes in the axial direction. The number of neutron energy group was increased from 7 to 18, and that of gamma energy group was increased from 3 to 7 groups respectively. As a result, HESTIA can accurately calculate the local neutron flux distribution and spectrum change within the fuel subassembly at the boundary between the fuel and reflector regions, which can not be fully simulated by MAGI. It was also confirmed that HESTIA can improve the power distribution in the driver fuel subassembly adjacent to radial reflector. As to thermo-hydraulic calculation, the porous body model was adopted to improve the calculation accuracy of coolant temperature. This model can take into account the detailed power distribution and the turbulent heat transfer in a fuel subassembly. It was found that the calculated value by HESTIA agreed well with that of the subchannel model. In order to verify the calculation accuracy of HESTIA, JOYO MK-II core characteristics calculation was conducted using HESTIA, and calculation results were compared with those of MAGI. The MAGI calculation results were already confirmed by the core performance test and post irradiation examination data. The comparison of both calculation code systems showed that the excess reactivity agreed within 0.01 % 
k/kk', the maximum neutron flux agreed within 3%, the ...
Maeda, Shigetaka; *;
JNC TN9400 2002-043, 43 Pages, 2002/07
JNC-TN9400-2002-043.pdf:1.71MB
It is important to precisely evaluate decay heat of the spent MOX fuel not only from the viewpoint of reactor safety concerning a decay heat removal at the reactor shut down, but also for the thermal design of the spent fuel storage and handling facility, In order to obtain the experimental data and to improve the accuracy of calculation, the decay heat of spent fuel subassemblies (Burn-up : 66 GWd/t) of the JOYO MK-II core was measured from the cooling time of 319 to 729 days. Measured decay heat of the spent fuel subassemblies was 351
16
1589W. The low flow rate measurement to make large temperature gadient between inlet and outlet of measurement system has allowed us to measure decay heat after cooling time of 500 days by less than 6% error(1
). The decay heat was calculated by "ORIGEN2" code using the JENDL-3.2 cross section library and the JNDC-V2 decay data and fission yield data library. Then it was compared with the measured value. The ratios between calculated and experimental value, C/Es, were approximately between 0.96 and 0.90. There exists a systematic discrepancy (68%) between calculation and measurement, which was larger than the experimental error (1 = 1.76.0%). Decay heat was generated from actinides and fission products (FP). Major heat sources in actinides are 
Cm, 
Pu and 
Am. However, nuclides except for Cm were not considered as the cause of systematic discrepancy because these decay heat was almost constant throughout this measurement. After cooling time of 100 days, 
Zr, Nb, 
Rh and 
Pr remained as major FP decay heat source. As a result, the discrepancy appears due to the decay heat calculation uncertainty of Cm and these FP, or measurement error and further investigation will be required.
Ariyoshi, Masahiko; Ishida, Koichi
JNC TN9400 2002-010, 25 Pages, 2002/03
JNC-TN9400-2002-010.pdf:0.67MB
This report is concerning the in-vessel in-service inspection (ISI) technology for the experimental fast reactor JOYO. The present ISI method in JOYO is not able to confirm the integrity of the core structure directly, expecting the visual inspection for the top of core assemblies from above the rotating plug. The purpose of this examination is to progress of the ISI method, and to confirm the integrity of the core structure directly. The core support plate is an important structure and it is selected for the object of in-vessel ISI. From the viewpoint of the influence on the plant, it is regarded impotant the method without all sodium draining, and the remote operation from above of the shielding plug. As a result, following technology is thought promising. (1)Under sodium inspection technique by means of ultrasonic method (it is able to apply in-vessel ISI without all sodium draining). (2)Nondestructive inspection technique by laser based ultrasonic method (it is superior in remote operating) (3)The local sodium discharge mechanism (it makes possible to apply laser based ultra sonic method for under sodium inspection) These technologies were investigated, examined, and the concept applied to ISI in JOYO core support plate was examined. Moreover, the problem when these were applied to in-vessel ISI in JOYO was picked up.
Ishida, Koichi; Maeda, Shigetaka; Saikawa, Takuya*; Masui, Tomohiko*
JNC TN9400 2002-005, 68 Pages, 2002/03
JNC-TN9400-2002-005.pdf:2.14MB
It is essential to evaluate the radiation damage of core structure materials used for core support plate and reactor vessel to maintain the safe operation of nuclear reactor plant. Therefore, surveillance tests for the irradiated specimen have been conducted in the experimental fast reactor JOYO to assure the integrity and to evaluate the life time. Neutron fluence and related spectral information are key palameters in evaluation of irradiation effects on the mechanical properties. They are usually predicted based on the calculation using the DORT two-dimensional transport code. In order to evaluate the calculation accuracy, the surveillance irradiation rigs (SVIRs) with dosimeter sets and gradient-monitor to monitor neutron fluences and temperatures were loaded several positions of the JOYO MK-II core. They were irradiated between 34
and 35 cycle. Based on the verification, the JOYO neutron field was precisely characterized and the calculated neutron flux at the positions of irradiated specimen and those of the core structure components need to be evaluated were corrected based on the experiments. As a result of this study, the following items are concluded: (1)The maximum fast neutron fluence (E
0.1Mev) on surveillance test specimen is determined as 2.07
10
n/cm
at 9th row of the core. (2)The neutron fluences at the positions of surveillance test specimen were higher than those of the core structure components. (3)For the core support plate which seems to be most critical for JOYO life time, the fast neutron fluence at present is 9.3810
n/cm and will reach 2.3110
n/cm at the end of life. The fast neutron fluence of reactor vessel is 3.1210
n/cm at present and will reach 4.8310n/cm at the end of life.
; ; Aoyama, Takafumi
JNC TN9400 2001-033, 45 Pages, 2001/01
JNC-TN9400-2001-033.pdf:7.55MB
Fission Products (FPs) were not considered in conventional fast reactor shielding analyses that were predominantly developed in clean core experiments like the JASPER program. However, in power reactors with high burn-up, the accumulation of FP affects the neutron balance so it cannot be neglected in the neutron flux calculation. In this study, the lumped group constants of FP nuclides were computed based on the JENDL-3.2 nuclear data library and these were compiled to the JSD-J2 set. Using the constants, the effect of the FP nuclides on shielding calculation was evaluated in the JOYO experimental fast reactor. Generation and depletion for nearly 880 FP nuclides can be computed with the ORIGEN2 burn-up calculation. The calculation uses the specification and material contents of the JOYO Mk-II driver as an example of fast reactor MOX fuel. About 99.8% of the total FP neutron absorption comes from 165 major nuclides. The cross section data for these nuclides are stored in the JENDL-3.2 library. The contributions of other FP nuclides were found to be negligible so the calculation used only these 165 FP nuclides. The lumped group constants for the FP nuclides were generated as follows. The 100 group infinite dilution cross section of each individua1 FP nuclide was computed with the NJOY-94 code. The energy group structure is the same as the JSD-J2 set and the scattering anisotropy is considered up to P3 components of Legendre expansion. Atomic number densities of FP nuclides generated from 
U, U, 
Pu and Pu were independently computed by ORIGEN2 as a function of fuel burn-up. The lumped FP constants were then obtained by averaging the infinite dilution cross sections with the atomic number densities based on the assumption that one fission produces one lumped FP. To verify the calculated lumped FP constants, the absorption microscopic cross section data were compared with the JFS-3-J3.2 group constants used for the fast reactor ...
; Aoyama, Takafumi; ; Shirakata, Keisho
JNC TN9400 2001-032, 57 Pages, 2001/01
JNC-TN9400-2001-032.pdf:1.53MB
This report describes the reactivity measurement method and its accuracy employed in ZPPR (Zero Power Physics Reactor) test facility at Argonne National Laboratory, USA. In ZPPR, reactivity measurements were conducted by means of MSM (Modified Neutron Source Multiplication) method using the fission reaction rates by 64 U fission chambers which were uniformly located in the core region. The MSM method can evaluate the subcritical reactivity considering the spatial distribution of neutron flux, although its basic equation is the one point approximation reactor dynamic theory. As each detector response varies depending on the core configuration, the uncertainty of the detector response affect directly the accuracy of measured reactivity. Therefore, the average value of many detectors had been conventionally used to reduce the statistical error in the measurerment. In this study, this method was modified to use the linear fitting between measured reaction rates and detector responses. Using this modified method, the uncertainty of each detector response affect the measured reactivity less than the simple averaging method. This new method was verified by applying to the ZPPR control rod worth measurements. It was found that the ratios of neutron source intensities and detector responses could be corrected accurately by the linear fitting in case of reactivity change without or with little fuel movement such as a unit control rod worth measurement.
Maeda, Shigetaka; Nose, Shoichi; *; Aoyama, Takafumi
JNC TN9400 2001-031, 39 Pages, 2001/01
Decay heat of the spent MOX fuel is important not only from the viewpoint of reactor safety concerning a decay heat removal at the reactor shut down, but also for the thermal design of the spent fuel storage and handling facility. In order to obtain the experimental data and to improve the accuracy of calculation, the decay heat of spent fuel subassemblies of the JOYO Mk-II core was measured. The burn-up was 66 GWd/t and the cooling time was between 40 and 150 days. Measured decay heat of the spent fuel subassemblies was approximately 1446 2466320W. The decay heat was calculated by "ORIGEN2" code and then it was compared with the measured value. In the "ORIGEN2" calculation, the JENDL-3.2 cross section library and the JNDC-V2 decay data library were used and fuel power calculated by the core management code system "MAGI" was used as a input. The ratios between calculated and experimental values, C/Es, were approximately between 0.94 and 0.90. The discrepancy between calculation and measurement was considered to be larger than the experimental error (1 = 1.73.0%) or the uncertainty of calculated FP decay heat (12%). It appears due to the uncertainty of actinides decay heat and that indicates cross sections of actinides and initial composition of actinides are important to evaluate decay heat accurately.
; Kitamura, Ryoichi; Aoyama, Takafumi
JNC TN9400 2000-071, 36 Pages, 2000/07
JNC-TN9400-2000-071.pdf:1.27MB
A three dimensional, continuous energy Monte Carlo code "MCNP" was applied to accurately evaluate the linear heat rate of mixed oxide (MOX) fuel used in a fast reactor. The test fuel pins to be analyzed were irradiated at the core center position in the experimental fast reactor JOYO MK-II core as the power-to-melt test, which was performed in order to optimize the thermal analysis method employed in the MOX fuel pin design. ln the calculation, the heterogeneous structure of the irradiation test subassembly (B5D-2) which accommodated test fuel pins inside was modeled precisely. The neutron flux distribution within the subassembly was calculated using MCNP with the neutron source which was based on the diffusion theory by the JOYO MK-II core management code system "MAGI" The linear heat rates of fuel pins were then obtained by multiplying the fission rate and fission energy of individual fissile nuclide. The gamma heating was calculated by MAGl considering the delayed fission gamma and it was included in the neutron heating obtained by MCNP. The accuracy of MCNP was verified by the post irradiation examination. using the fission product yields (
Nd) measured for the melted part of test fuel pins, the linear heat rates were evaluated and compared with the MCNP results. The average ratios of MCNP calculations to the experimental values were 0.955 0.020, indicating MCNP can precisely calculate the linear heat rates. By correcting the MCNP calculation results with the C/E values obtained above, the linear heat rates of test fuel pins were evaluated to be 620 - 685 W/cm at the core center height.
Harano, Hideki; Nose, Shoichi;
JNC TN9400 2000-076, 34 Pages, 2000/05
JNC-TN9400-2000-076.pdf:0.67MB
lmmediate detection of fuel failure and subsequent precise identification of failed fuel assembiies are extremely important and indispensable for fast reactors from the viewpoint of their safety and reliability as well as the improvement of plant availability. ln order to develop the failed fuel detection and location (FFDL) technology, laser resonance ionization spectrometry (RIMS) has been proposed to be applied to the trace analysis of krypton and xenon contained in cover gas, Various promising features have been reported including the results which suggest the feasibility of the method to the on-power real-time monitoring, through the fundamental study using the RIMS device at the nuclear engineering research laboratory (NERL) of the university of Tokyo. Based on the information obtained above, we are developing a new laser FFDL system using RIMS which is planned to be introduced onto the fast experimental reactor JOY0. By the use of the system at the JOY0, isotope analysis can be performed with high sensitivity for not only radioactive but also stable elements in fission product (FP) and tag gas in the cover gas. This permits the improvement of irradiation technology and the immediate identification of failed fuel assemblies. For instance, it is possible to identify burst samples in the breach test of fuel cladding materials during irradiation. From the isotopic composition of the FP nuclides, the burnup of failed fuel can be estimated which allows the preliminary focusing in the FFDL. ln this paper, we review the fundamental study using the RIMS device at NERL and report the basic design of the laser FFDL system for the JOY0.
JNC TN9400 2000-062, 61 Pages, 2000/03
JNC-TN9400-2000-062.pdf:1.78MB
lt has been studied for safety plant test to demonstrate inherent safety of FBR core in JOYO. As a part of this study investigation of feedback reactivity was studied. And analysis code which can estimate the reactvity of core-bowing has been prepared. In post irradiation examination, remain bowing displacement of sub-assembly can only be measured, so development of measuring system of displacement of sub-assembly under operation is expected to improve estimation of core-bowing reactivity. lt is supposed that an on-line measurement system of displacement of the top of sub-assembly using super sonic sensor under operation is an effective means. ln this report, development of measurement of displacement of the top of sub-assembly using super sonic sensor was settled as follows. (1)Characteristic test of heat-resistant super sonic sensor (2)Design of adjustment device of super sonic sensor (3)Examination of in-water test for estimation of temperature fluctuation
; *
JNC TN9400 2000-001, 42 Pages, 1999/12
JNC-TN9400-2000-001.pdf:1.51MB
Neutron flux monitoring can be an effective method for the early detection of reactivity anomalies in a FBR core. This development required that the normal range of reactivity fluctuations at power should be well characterized. An analysis of Reactor noise showed that the low frequency power fluctuation was influenced by the coolant temperature fluctuation, and that the high frequency power fluctuation was caused by control rod vibration. But spectral resolution was not adequate to determine the normal range of power fluctuation quantitatively. Also, a transfer function for the coolant temperature to reactor power ratio was modeled with the time constants of the thermal expansion of the core support plate, coolant temperature measurements, and so on. This was necessary to clearly understand the cause of the normal power fluctuations. The calculated values of simulated reactor power were compared with typical power ratio data from JOYO and the comparison was good in the low frequency range. Hence, it is clear that the dynamic characteristics of reactor power are caused by coolant temperature fluctuations. The high frequency power fluctuation, which is caused by control rod vibration, can be separated from the normal reactor power fluctuation. This analysis has shown that the normal range of reactor power fluctuations can be quantitatively determined accurately, and that neutron flux monitoring can be applied to detect reactivity anomalies early in a FBR core.
Sumino, Kozo; ; Maeda, Yukimoto; Aoyama, Takafumi; ;
PNC TN9430 98-006, 141 Pages, 1998/09
None
Sumino, Kozo; ; Maeda, Yukimoto; Aoyama, Takafumi; ;
PNC TN9430 98-003, 139 Pages, 1998/06
None