; *; Sakurai, Koji*; *; *; *
JNC-TN8400 2000-032, 98 Pages, 2000/12
Concerning the preparation of high U solution for the crystallization process and the application of UO powder dissolution to that, the effects of final U concentration, dissolution temperature, nitric acid concentration and powder size on the dissolution of UO powder in the nitric acid where the final U concentration was 800g/L were investigated. The experimental results showed that the solubility of UO decreased with the increase of final UO concentration and powder size, and with the decrease of dissolution temperature and nitric acid concentration. It was also confirmed that in the condition where the final U concentration was sufficiently lower than the solubility of U,,UO dissolution behavior in the high U solution could be estimated with the equation based on the fragmentation model which we had already reported. Based on these experimental results, the dissolution behavior of irradiated MOX fuel in high U solution was estimated and the possibility of supplying high U solution to the crystallization process was discussed. In the preparation of high U solution for the crystallization process, it was estimated that the present dissolution process (dissolution for fuel pieces of about 3cm long) needed a lot of time to obtain a high dissolution yield, but it was shorted drastically by the pulverization of fuel pieces. The burst of off-gas at the early in the dissolution of fuel powder seems to be avoidable with setting the appropriate dissolution condition, and it is important to optimize the dissolution condition with considering the capacity of off-gas treatment process.
JNC-TN8410 2000-015, 7 Pages, 2000/10
Some falsification has been detected in the results of quality control data relating to the diameter of samples of pellets produced in the BNFL's MOX Demonstration Facility (MDF) on September 1999. This document is the outlines of inspection procedure for the MONJU fuel pellet in plutonium fuel center of JNC.
; ; ; Matsumoto, Shinichiro
JNC-TN9410 2000-009, 65 Pages, 2000/09
In order to evaluate irradiation behavior of(U, Pu) C and (U, Pu) N fuel using fast reactor, (U, Pu) C and (U, Pu) N fuel pins were irradiated in JOYO for the fist time in Japan. In this study, one (U, Pu) C fuel pin and two (U, Pu) N fuel pins were irradiated to maximum burn up about 40GWd/t. Post irradiation examination of (U, Pu) C and (U, Pu) N fuel pins started in Fuel Monitoring Facility (FMF) at JNC from October 1999, and it ended in March, 2000. The results of non-destructive post irradiation examination reported in this document. Main results are shown in the following. (1)The soundness of all (U,Pu) C and (U,Pu) N fuel pins were confirmed from the non-destructive examination result. (2)The fuel stack elongation of (U,Pu) C and (U,Pu) N is bigger than it of the MOX fuel for fast reactor. (3)The singular behavior from the gamma ray scanning measurement in the stack area was not confirmed. The migration of Cs137 to lower insulator pellet and outside of the pellet was confirmed in (U,Pu) N B9NO2 pin. In (U,Pu) C fuel, the migration of Cs137 was not confirmed. (4)In (U,Pu) C B9CO1 pin and (U,Pu) N B9NO2 pin in which the gap width was small, diameter of cladding increase around 50 m in the stack area which originates for FCMI was confirmed. In (U,Pu) N B9NO1 pin in which the gap width was wide, the ovality which originates from the relocation of the pellet was confirmed. (5)Fission gas release rate of (U,Pu) N were 3.3% and 5.2%, and the low value compared to the MOX fuel was shown.
Yamanaka, Shinsuke*; Uno, Masayoshi*; Kurosaki, Ken*; ;
JNC-TY9400 2000-011, 41 Pages, 2000/03
no abstracts in English
JNC-TN9400 2000-041, 29 Pages, 2000/03
Irradiation behavior and performance models were investigated in order to apply for nitride fuel options in feasibility study on fast breeder reactor and related recycle systems. (1)MechanicaI design of nitride fuel pin: The behaviors of fission gas release (increase of internal Pressure) and fuel-to-cladding chemical interaction (decrease of cladding thickness) are needed to evaluate cumulative damage fraction in case of fuel pin mechanical design. The behaviors of fission gas release and fuel-to-cladding chemical interaction were investigated from the past studies up to high burnuP, since the lower fission gas release in nitride fuel than in oxide fuel could contribute to reduce the plenum volume and result in the shortening of fuel Pin length. (2)Fuel pin smear density: The higher fuel smear density is preferred for the higher fissile density to improve the core characteristic. The behaviors of fuel pellet swelling were investigated from the past studies up to higher burnup, since the larger fuel pellet swelling in nitride fuel than in oxide fuel would restrict high burunp capability due to fuel-cladding mechanical interaction. (3)Compatibility of nitride fuel with high Temperature water: Compatibility of nitride fuel with high temperature water were investigated from the past studies to contribute water cooled fast breeder reactor options.
*; 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 ...
JNC-TN9400 2000-029, 87 Pages, 1999/11
The second Power-To-Melt (PTM) test, PTM-2, was performed in the experimental fast reactor "JOYO". AIl of the twenty-four fuel pins of the irradiation vehicle, B5D-2, for the PTM-2 test, were provided for post-irradiation-experiment (PIE) to evaluate the PTM values. ln this study, the PIE technique for PTM test was established and the PTM results were evaluated. The findings are as follows: (1) The maximum fuel-melting ratio on the transverse section was 10.7%, and was within the limit of fuel-melting in this PTM test enough. Unexpected fuel-melting amount to a ratio of 11.8% was found at 24 mm below the peak power elevation in a test fuel pin, lt is possible that this arose from secondary fuel-melting. (2) Combination of metallographical observation with X-ray microanalysis of plutonium distribution was very effective for the identification of once-molten fuel zone. (3) The PTM evaluation suggested that dependence of the PTM on the fuel pellet density was stronger than that of previous foreign PTM tests, while the dependence on the pellet-cladding gap and the oxygen-to-metal ratio was indistinctly. The dependence on the cladding temperature and the fill gas composition was not shown as well.
; ; ; ; ;
JNC-TN8410 98-007, 201 Pages, 1998/11
PNC-TN8470 97-003, 55 Pages, 1997/11
no abstracts in English
; ; 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 BC 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 ...
; ; *; ; ; ;
PNC-TN8410 97-045, 28 Pages, 1997/03
PNC-TJ8005 97-001, 122 Pages, 1997/03
no abstracts in English
; ; ; ; ; Kajitani, Yukio;
PNC-TN9410 97-015, 382 Pages, 1996/12
For the purpose of developing the future nuclear fuel recycle system, the design study of the advanced nuclear fuel recycle system is being conducted. This report describes intermediate accomplishments in the conceptual system study of the advanced nuclear fuel recycle system. Fundamental concepts of this system is the recycle system using molten salt which intend to break through the conventional concepts of purex and pellet fuel system. Contents of studies in this period are as follows, (1)feasibility study of the process by Cd-cathode for nitride fuel (2)application study for the molten salt of low melting point (AlCl+organic salt)(3)research for decladding (advantage of decladding by heat treatment)(4)behavior of FPs in electrorefinning (behavior of iodine and volatile FP chlorides, FPs behavior in chlorination) (5)criticaliy analysis in electrorefiner (6)drawing of off-gas flow diagram (7)drawing of process machinery concept (cathode processor, vibration packing) (8)evaluation for the amounts of the high level radioactive wastes (9)quality of the recycle fuels (FPs contamination of recycle fuel) (10)conceptual study of in-cell handling system (11)meaning of the advanced nuclear fuelrecycle system. The conceptual system study will be completed in describing concepts of the system and discussing issues for the developments.
Kato, Masato; ;
PNC-TN8410 96-247, 97 Pages, 1996/08
; ; ;
PNC-TN8440 96-034, 109 Pages, 1996/07
PNC-TN8410 96-218, 216 Pages, 1996/07
; ; ; ;
PNC-TN8410 96-214, 36 Pages, 1996/07
; ; ; ; ; ;
PNC-TN8410 96-198, 224 Pages, 1996/06
Arai, Yasuo; Iwai, Takashi; ; Okamoto, Yoshihiro; Shiozawa, Kenichi;
JAERI-Research 96-009, 17 Pages, 1996/02
no abstracts in English
Nippon Kikai Gakkai-Shi, 98(922), p.759 - 761, 1995/09
no abstracts in English