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Soroi, Masatoshi; Miyakawa, Shunichi
PNC TN9410 98-050, 57 Pages, 1998/05
This report describes the present situation and problems with the development of the flow control irradiation facility (FLORA). The purpose of FLORA is to run the cladding breach (RTCB) irradiation test under loss of flow conditions in the experimental fast reactor "JOYO". FLORA is a facility like FPTF (Fuel Performance Test Facility) plus BFTF (Breached Fuel Test Facility) in EBR-II, USA. The technical feature of FLORA is its annular linear induction pump (A-LIP), which was developed in response to a need identified through the experiences in the mechanical flow control of FPTF. We have already designed the basic system facility of FLORA for the JOYO MK-II core. However, to put FLORA to practical use in the future, we have to confirm the stability of the JOYO MK-III core condition, solve problems and improve the design. The main results and problems of the development of FLORA are as follows; (1)The results of the development: (a)The neutron detector in FLORA can detect the delayed neutron which is emitted from failed fuel. (b)Out-of-pile A-LIP tests in sodium conditions has been completed. (The length of the tested A-LIP is half the actual size.) Out-of-pile test results showed that the A-LIP achieved a 300
/min flow rate and 265kPa pressure in 550
C sodium. This pump performance satisfied the FLORA requirements. (c)By controlling the sodium flow rate from 40 to 100% using the A-LIP, we can control the fuel cladding temperature satisfactorily. (2)The problems: (a)In the development of the process detector, it is necessary to miniaturize the neutron detector and test the effect of neutron irradiation and high temperatures on the permanent magnet in the flow meter. (b)The problem which is left about A-LIP is its influence on neutron irradiation. For this purpose, we have to irradiate a small size A-LIP and test its characteristics and electric isolation. (c)To get more accurate results concerning the efficiency of the A-LIP, we have to ...
; *; Miyakawa, Shunichi; *;
PNC TN9410 98-035, 60 Pages, 1998/03
This report describes the development activities for the fabrication of the Thermal Expansion Difference irradiation temperature monitor (TED) at the Oarai Engineering Center (OEC)/PNC. TED is used for various irradiation tests in the experimental fast reactor JOYO. TED is the most accurate off-line temperature monitor used for irradiation examination. The TED is composed of a metallic sphere lid and either a stainless steel or nickel alloy container. Once the container is filled with sodium, the metallic sphere lid is sealed by using a resistance weld. This capsule is then loaded into a reactor. Once a TED is loaded into the JOYO reactor, the sodium inside the metallic container increases as a result of thermal expansion. The TED identifies the peak irradiation temperature of the reactor based on a formula correlating temperature to increment values. This formula is established specifically for the particular TED being used during a calibration process performed when the TED is fabricated. Initially the TED was developed by Argonne National Laboratory (ANL) in the United States, and was imported by PNC for use in the JOYO reactor. In 1992 PNC decided to fabricate TED domestically in order to ensure the stability of future supplies. Based on technical information provided by ANL, PNC began fabrication of a TED on an experimental basis. In addition, PNC endeavored to make the domestically produced TED more efficient. This involved improving the techniques used in the sodium filling and the metallic sphere welding processes. These quality control efforts led to PNC's development of processes enabling the capsules to be filled with sodium to nearly 100%. As a result, the accuracy of the temperature dispersion in the out-pile calibration test was improved from 
10C to 5. In 1996 the new domestically fabricated TED was attached to a JOYO irradiation rig. In March of 1997, irradiation of the rig was started on the 30th duty cycle operation, .
Isozaki, Kazunori;
PNC TN9410 97-069, 134 Pages, 1997/07
Hyougo-ken southern earthquake broke out in 1997/01/17. The Atomic Energy Safety Co㎜ission considered reasonable of the design guide for seismic design. And the Science and Technology Agency(STA) required reevaluation of atomic power facilities built by old design guide according to the new seismic design guide. JOYO obtained the construction license in 1970/02. Heat transport system and buildings of JOYO was reevaluated by the new seismic design guide for the MK-III project. So, JOYO was not required reevaluation by STA. But, this evaluation of MK-III was limited to reconstruction area, and the seismic design was reevaluated extensively to confirm earthquake proof characteristics. The structural integrity of buildings and equipments was confirmed by the result of reevaluation by the new seismic design guide. The analysis model conditions were established according to the 1987 and 1991 version of JEAG. This was done by ground investigation result and buildings vaibration test. It was made clear that the analysis model conditions were reasonable and conservative from a technical view point.
Miyakawa, Shunichi; ; Soga, Tomonori
PNC TN9410 97-068, 113 Pages, 1997/07
Since the first control rod design for the Joyo Mk-II core (about twenty years ago), there have been several challenging improvements; for example, a helium venting mechanism and a flow induced vibration prevention mechanism. Forty-four control rods with these various modifications have been fabricated. To date, thirty-four have been irradiated and the sixteen have been examined, This experience and effort has produced fruitful results: (1)Efficiency and reliability of the diving-bell type Helium venting mechanism (2)Efficiency of the flow induced vibration prevention mechanism (3)Efficiency of the improvement for scram damping mechanism (4)Clarification of absorvber-pellet-cladding-mechanical-interaction (ACMI)phenomena and preventive methods The fourth result listed above has been a subject of investigation for fifteen years in several countries, that is a main phenomena to dominate control rod life time. The results of this investigation of ACMI in absorber elements are summarized below: (a)In five of Joyo Mk-II control rods, cladding cracks were found in fifteen of the elements. These cracks were caused by a acceleration ACMI, due to B
C fragments relocation. They occurred over a wide burnup range from 5E+26 Cap./m
to 45E+26Cap./m in a nearly typical provability distribution. The cladding cracked because of its low ductility (approximately 1/4 lower than the uniform elongation of usual tensile testing for irradiated 316SS cladding) due to neutron irradiation and the ultra slow ACMI induced strain rate. (b)In this case the crack growth rate is extremely slow and the ACMI induced cracking in absorber elements do not influence either the reactor or plant operations. It is on this basis that a strict limitation to avoid the cladding crack is not necessary. According1y, it is suggested that a realistic design standard should consider the ACMI phenomena and the burnup limit be based on the nominal base calculation for average plastic strain use ...
Isozaki, Kazunori;
PNC TN9410 97-062, 169 Pages, 1997/05
Sodium leak accident of MONJU was caused high cycles fatigue damage of thermometer well by flow-induced vibration. It was due to the sy㎜etric vortex shedding which was occurred rear flow of thermometer well. So, Thermometer wells installed in primary and secondary heat transport systems of JOYO were evaluated of flow-induced vibration. Evaluation of flow-induced vibration of thermometer well was done checking of flow-induced vibration base on authorized design report for JOYO, evaluation of summary flow-induced vibration by natural frequency of thermometer well in sodium as cantilever models, and evaluation based on small velocity rule of ASME Code Section III Appendix N-1300. By this result, thermometer wells (12B piping of secondary cooling system) were not sattisfied requirement to avoid flow-induced vibration by small velocity rule. Therfore, Detailed vibration characteristic analysis, water flow-induced vibration test, dumping test and evaluation of structural integrity were carried out. These results, vibration amplitude of well on the tip was 0.13mm (vibration non-dimensional amplitude of 0.015) and peak stress of 2.9kg/mm
is occurred. Thermometer wells (12B piping of secondary cooling system) which occurred peak stress by flow vibration was confirmed enough to satisfy 5.3kg/mm2 of design fatigue limit.
Ito, Hideaki; ; ; Okubo, Toshiyuki
PNC TN9410 96-298, 177 Pages, 1996/11
The fuel handling facility in "JOYO" must maintain an argon atmosphere and be gas tight; this prevents the oxidation of sodium adhering to a fuel assembly and leakage of radioactive gases. Periodic leak testing of the double O-ring gas seal had been performed at increasing pressure to assure its specified leak tightness. The problem with this method was that it took a long time to obtain an accurate measurement. The leak testing methods for the fuel handling facility, the reactor containment vessel, and other vessels were all reexamined. As a consequence, it was determined that alternative devices and methods for improving the leak rate measurements should be studied. Four methods of leak testing were evaluated; the present increasing pressure method, helium leak testing, decreasing pressure method, and a liquid nitrogen decreasing pressure method. A new automatic leak measurement device was used in these evaluations. The results of the utilization and limitations of the four methods of leak testing are summarized as follows. (1) The decreasing pressure leak testing method was efficient with regard to accuracy and stability for use in the fuel handling facility. (2) The automatic leak measurement device used a statistical calculation to measure the leak rate stability and it met the specified measurement requirements. (3) The leak rate measuring time was reduced by half with this new device and it could also simultaneously examine other objects.
Aoyama, Takafumi;
PNC TN9420 96-058, 27 Pages, 1996/10
This report introduces the nuclear instrumentation system and the major radiation measurement techniques used in the Experimental Fast Reactor "JOYO". In the introduction of the nuclear instrumentation system, system function and role as reactor plant equipment, specifications and characteristics of neutron detectors, and layout of the system are described. Reactor dosimetry was used to evaluate neutron dose and their spectra for various irradiation tests and surveilance tests performed in JOYO. The multiple-foil activation method which is currently used and the Helium Accumulation fluence Monitor (HAFM) under development are described. The fuel failure detection (FFD) and the failed fuel detection and location (FFDL) systems in which radiation measurement plays a key role are introduced. It was shown some of the major experimental results obtained from a series of fuel failure simulation tests performed in JOYO. Finally, as a new radiation measurement technique, the Plastic Scintillation Fiber (PSF) is described which is a position sensitive radiation detector that can detect the radiation dose rate at the relevant position in the fiber. The PSF is used to upgrade the gamma-ray distribution measurement to accurately evaluate the Corrosion Products (CPs) behavior in the JOYO primary coolant system.
; ; ; *
PNC TN9410 96-269, 24 Pages, 1996/09
As the factor which affects the reactivity coefficient, there is a thing by the change of the flow characteristic of the core internal structure by the mechanical deformation. This displacement of core support-plate in the JOYO reactor vessel was analyzed by using "FINAS" (Finite Element Nonlinear Structural Analysis System). This displacement is a result of the coolant loading difference between the shutdown and full operational primary flow rates. The displacement of core support-plate in the JOYO reactor vessel was analyzed by using "FINAS" (Finite Element Nonlinear Structural Analysis System). This displacement is a result of the coolant loading difference between the shutdown and full operational primary flow rates. At first, the deformation of the core support was required according to the two-dimensional axisymmetric model. These displacements were used as input in the three-dimensional models of excluding the support rib. As a result, the displacement of the upper core support-plate would be about +0.39mm in the center and the maximum would be about +0.43mm in the fifth row. Because of the displacement of the support-plate was suppressed by the support rib. The whole core support structure was displaced in the upward direction. And, core support-plate of the upper and lower was spread largest about 0.03mm in the tenth row by the high-pressure plenum coolant loading.
Myochin, Munetaka; Kosugi, Kazumasa; Wada, Yukio; Yamada, Kazuo; Seimiya, Hiroshi; Ishikawa, Hirohisa
PNC TN8410 96-071, 86 Pages, 1996/03
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*; *; *; Sato, Wakaei*; *; Sanda, Toshio*
PNC TN9410 95-214, 199 Pages, 1995/08
In order to improve the design method and accuracy of large fast breeder cores, extensive work has been performed to accumulate and evaluate many kinds of results of fast reactor physics experiments and analyses. As a part of efforts to develop a standard data base for LMFBR core nuclear design, the present report evaluates the physical consistency of JUPITER experimental analysis, especially concentrating on criticality. Here, the judgment of consistency is based on not only the deviation degree of C/E values from unity, but also various viewpoints such as the comparison with other cores or other nuclear characteristics by sensitivity analysis, the effect of changing nuclear data library, the analysis of FCA and JOYO which have completely different source of data from JUPITER, and the use of the Monte Carlo method as an analytical reference. (1)The C/E values of JUPITER criticality are slightly underestimated in the range of 0.993-0.999, using the JFS-3-J2 (1989) group constant set based on JENDL-2 and three-dimensional XYZ transport theory with the most detailed analytical model. There is an obvious dependency of C/Es on reactor core concepts with homogeneous or heterogeneous structure, the main cause of which is considered to be the effect of internal blanket existence and cross-section errors of JFS-3-J2, judged from sensitivity analysis. (2)The latest analytical method and model based on three-dimensional XYZ transport theory has sufficient ability to predict the relative changes of JUPITER criticality caused by the effect of reactor core size, CRP sodium channel, control rod and internal blankets. (3)The analytical error of JUPITER criticality was evaluated as approximately 0.3%dk and this seems reasonable, because the results of Monte Carlo analysis for ZPPR-9 criticality were almost identical with those of our standard analytical method. (4)The analytical results based on the latest JENDL-3.2 library were very close to those of JENDL-2 results, ...
; *; *; Tanimoto, Kenichi; Enokido, Yuji
PNC TN9410 92-208, 68 Pages, 1992/07
The plan of the gelogic disposal of high level waste that need to do estimate the effect of radiation in near field surrouding waste. We have applied "JOYO" spent fuel storage pool as irradiation field and investigated the effect of gamma radiation about quality of groundwater, because we have to obtained a basic data related geologic dispose under irradiaton condition. The experiment was applied artificial brine for simulated groundwater. The same samples were located in "JOYO" spent fuel storage pool without gamma radiation and other effects were estimated. The samples were also observed the variation of quality of artifical brine every fixed time after irradiation and were estimated the effect as the function of time, gamma irradiation were carried out from 24hours (1.0
10
1.310Gy) to 1440 hours (4.410
6.810Gy). The results indicate the following. (1)The change of pH, conductivity and ion concentrations in artifical brine could not be observed in the samples before and after irradiation. (2)Eh of the samples was 241mV before irradiation, but it decreased 156mV after irradiaton for 1440 hours. Eh tend to decrease by increase of the absorption dose. (3)Do of the samples before and after irradiation for 1440 hours were 20.76 and 5930 
g/, respectively. Do tend to increases by increase of the absorption dose. (4)Before and after irradiation test for 480 hours, nitric ion was detected 2.9 and 105ppm, respectively, In no gamma irradiaton test, nitric ion was detdcted 4.0 and 5.6ppm, respectively. For 1440 hours, nitric ion was detected 15ppm after irradiation and 11ppm after rest without gamma irradiation. (5)pH, Eh, Do, conductivity and all ion concentrations in artifical brine have no the variation as the function ot time within fixed time (about 4hours) after irradiation. These results suggest that oxygen which were generated by the gamma radiolysis of water was incrased Do, ...
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.
; Kawasaki, Hirotsugu; ; ; Tsukimori, Kazuyuki; ; Isozaki, Kazunori
PNC TN9410 92-131, 90 Pages, 1992/05
This report describes the summary of 'A validation Test Planning of Piping Bellows Expansion Joint in Experimental FBR "JOYO" '. The planning followed 'A Feasibility Study of Piping Bellows Expansion Joint' which had been completed at oarai Engineering Center of PNC. The planning of validation test is one activity of the First Sub-Committee of the PROFIT Committee. Under the First Sub-Committee, a working group was organized for constructing the varidation test program. The working group had considered meanings/objectives, methods, schedules, expected results and expenses of validation test using experimental FBR 'JOYO'. The output of the working group presented in this repot was reported to and deliberated in the First Sub-Committee, the PROFIT Committee and a Technical Stearing Comittee in Oarai Engineering Center. The last choice was made by the Technical Stearing Comittee to push forward with the plan.
; *; ; ; ; Matsumoto, Mitsuo
PNC TN8410 92-187, 21 Pages, 1992/05
Computer code for fuel temperature evaluation of LMFBR at early burnup has been developed. On fule temperature evaluation, especially at early burnup, the fuel restructuring model and the gap conductance medel are important. These models which are installed in the temperature evaluation code, were verified based on the results of irradiation tests using the foreign fast reactor and "JOYO". This paper describes the essential parts of the models and the functions of the code (DIRAD) which is used for the fuel temperature evaluation at early burnup of LMFBR such as the prototype fast breeder reactor "MONJU".
