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JNC TN9440 2000-005, 164 Pages, 2000/06
JNC-TN9440-2000-005.pdf:4.51MB
This report summarizes the operating and irradiation data of the experimental reactor "JOYO" 34th cycle, and estimates the 35th cycle irradiation condition. Irradiation tests in the 34th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup perfomance 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)Absorber Materials Irradiation Rig (AMIR-6) (a)Run to absorber pin's cladding breach (4)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (5)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large reactor (6)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (7)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confirmation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (in collaboration with universities) The maximum burnup driver assembly "PFD537" reached 68,500MWd/t(pin average).
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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.
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PNC TN9410 97-064, 42 Pages, 1997/06
In our study on a hyblid fast reactor system, we have investigated it from the view point of transmutation ability of trans-uranium (TRU) nuclide making the most effective use of special features (controllability, hard neutron spectrum) of the system. It is proved that a proton beam is superior in generation of neutrons compared with an electron beam. Therefore a proton accelerator using spallation reaction with a target nucleous has an advantage to transmutation of TRU than an electron one. A fast reactor is expected to primarily have a merit that the reactor can be operated for a long temrm without employment of highly enriched plutonium fuel by using external neutron source such as the proton accelerator. Namely, the system has a desirable characteristic of being possible to self-sustained fissile plutonium. Consequently in the present report, core characteristics of the system were roughly studied by analyses using 2D-BURN code. The possibility of self-sustained fuel was investigated from the burnup and neutronic calculation in a cylindrical core with 300w/cc of power density without considering a target material region for the accelerator. For a reference core of which the height and the radius are both 100cm, there is a fair prospect that a long term reactor operation is possible with subsequent refueling of natural uranium, if the medium enriched (around 10wt%) uranium or plutonium fuels are fully loaded in the initial core. More precise analyses will be planed in a later fiscal year.
; Nagaoki, Yoshihiro
PNC TN9410 97-022, 34 Pages, 1997/02
The core averaged burn-up reactivity has been measured and calculated for the Joyo MK-II core. In order to evaluate the relationship between the calculational error of burn-up reactivity and the nuclear data or calculated neutron flux, the burn-up reactivity for an individual fuel subassembly(S/A) must be measured. So the burn-up reactivity measurement test was conducted on the MK-II core. The burn-up reactivity for a driver fuel S/A was measured as a substitution reactivity worth between two S/As at different burn-ups. In the test a fuel S/A with a burn-up of 1 GWd/t was substituted by two S/As with 37 and 62 GWd/t, respectively. The substitutions were carried out at the core center(row 0), middle of the fuel region(row 2)and the border region of the fuel and reflector(row 4). The calculated burn-up reactivity worth performed by the core management code system "MAGI" was compared with the measured value. The results obtained were as follows: (1)Measured substitution reactivity worth(at row 0) between 1 and 37 GWd/t fuel S/A was -0.19%
k/kk' and that between 1 and 62 GWd/t was -0.28%
k/kk'. (2)Relative distribution of the reactivity worth between 1 and 37 GWd/t agreed with that between 1 and 62 GWd/t. The relative value normalized at core center was 0.67 for the row 2 and 0.28 for the row 4. (3)The C/E value was 1.03
1.05 for the substitution between 1 and 37 GWd/t and 0.940.95 between 1 and 62 GWd/t at the row 0 and 2. It was clear that the C/E values at the row 4 are higher than those at the row 0 and 2 by 57%. An analysis of the burn-up dependency on the C/E value of the burn-up reactivity worth is being performed in detail. Presents, the PIE of the fuel S/A used for the measurement is under way.
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PNC TN9410 97-013, 111 Pages, 1997/01
Since the 54-rod cluster high-burnup fuel assemblies are planing to be loaded in Fugen, it must be confirmed that the mechanical integrity of the assemblies will be retained during the dwelling period in the reactor core. In the integrity verification, the confirmation that fretting wear, which occurs on fuel cladding surface at the contact with the spacer ring-elements, will not exceed the design margin is important. Accurate measurements of the flow-induced vibration characteristics under the hydraulic condition of coolant simulating the reactor core, especially measurements of the vibration amplitude, is necessary because the vibration amplitude directly affects the fretting wear depth. The flow-induced vibration measurements of the 54-rod cluster high-burnup fuels in which accelerometers were installed, were carried out under the various hydraulic conditions in the Component-Test-Loop (CTL). The results of the measurements are discribed in this papers. From the frequency analysis, the characteristic frequency of the fuel was observed around 105 Hz and 160 Hz. This frequency approximately coincided with that estimated by the fretting wear analysis code. The amplitude of flow-induced vibration was increased with increase in total flow rate and steam quality. Though these tendencies coincided with the results calculated by the analysis code, the amplitude measured at the region of low flow rate tended to be large compared with the calculated values. It was confirmed that this difference can be reduced on the safety side by the modification of the equation in the analysis code. The Paidoussis equation is divided into two terms in this modification, in which one term depending on total flow rate and the other term depending on steam quality, and proper coefficients are determined for each term. Though the amplitudes of flow-induced vibration for this fuel were larger than for either of the 28-rod cluster fuel of Fugen and 36-rod cluster fuel of ATR demonstration ...
; ; Yamada, Takeshi; ; ; ; Kaito, Yasuaki; ; Kotaka, Yoshinori; ; et al.
PNC TN2410 96-005, 339 Pages, 1996/03
Construction of the 'Monju' fuel handling systems was completed in April, 1992. From March 1991 to August 1992, pre-commissioning tests were carried out. In December 1992, all the systems of Monju were transfered to PNC, and commissioning tests and reactor physics tests, were started. For the first time, during these physics tests, the fuel handling systems were operated for one of the commissioning tests 'Loading to Criticality', without significant problems. 168 fuel sub-assemblies were loaded into the core and the first criticality was achieved on 5th April 1994. The fuel handling systems continued in operation for the 'Loading to Full Size of the Core', power distribution test and for cleaning discharged dummy sub-assemblies. To keep these fuel handling systems working somothly and satisfactorily annual maintenance has been carried out since 1992. This paper describes the operation and maintenance experience of fuel handling systems after the pre-commissioning tests and future study items for system reliability improvement.
*; *; Tone, Tatsuzo
JAERI-M 6876, 25 Pages, 1977/01
no abstracts in English
Yokota, Satoru; Hatanaka, Akira; Fujimori, Masahito; Shimoyamada, Tetsuya; Nakamura, Yoshinobu
no journal, ,
Three batch-type dissolvers in the Tokai Reprocessing Plant are a device for dissolving spent fuel. The dissolver is composed of one slab and two barrels (stainless steel 310s). Install a fuel basket (stainless steel 304L) in the barrel and accept the sheared spent fuel to dissolve it. The insoluble fuel cladding is taken out of the barrels with the basket. The dissolution time of operation for one batch is approximately 10 hours. During dissolution operation, nitric acid was added to the dissolver into the spent fuel in the basket with water. The solution was heated with steam. Corrosion failure has occurred in the past because the dissolver is exposed to a high corrosive environment (high temperature, high acid concentration). Therefore, we carry out the periodical wall thickness measurement of the barrel by the remote control. On the other hand, the wall thickness measurement of the fuel basket was carried out only once by destructive measurement at the time of renewal in 1999. The details of the corrosion tendency of the fuel basket are unknown, and it is urgent to establish a non-destructive measurement method by remote handling. Therefore, we examined the method of wall thickness measurement of the fuel basket and established the measuring technique.