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Akie, Hiroshi; Nakamura, Takehiko
Progress in Nuclear Energy, 38(3-4), p.363 - 370, 2001/02
Times Cited Count：6 Percentile：44.09(Nuclear Science & Technology)no abstracts in English
; ; ; Matsumoto, Shinichiro
JNC TN9410 2000-009, 65 Pages, 2000/09
JNC-TN9410-2000-009.pdf:4.36MB
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.
*
JNC TN9440 2000-008, 79 Pages, 2000/08
JNC-TN9440-2000-008.pdf:2.33MB
This report summarizes the operating and irradiatlon data of the experimental reactor "JOYO" 35th cycle. Irradiation tests in the 35th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup performance 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)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (4)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large scale reactor (5)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (6)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confimation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (based on a contract with universities) The maximum burnup driver assembly "PFD253" reached 67,600 MWd/t (pin average).
; ; Sakamoto, Naoki; *; Akasaka, Naoaki;
JNC TN9400 2000-095, 110 Pages, 2000/07
JNC-TN9400-2000-095.pdf:13.57MB
The effects of high fluence irradiation and swelling on the transient burst properties of austenitic steel fuel claddings; PNC316 and 15Cr-20Ni stcel, which were irradiated as the MONJU type fuel assemblies (MFA-1&MFA-2) in the FFTF reactor, were investigated. The temperature-transient-to-burst tests were conducted on a total of eight irradiation conditions. Fractographic examination and TEM observation were performed in order to evaluate the effect of high dose irradiation on the transient burst property and the relation between failure mechanism and microstructural change during rapid (ramp) heating. The results of the PIE showed that there was no significant effect of irradiation on the transient burst properties of these fuel claddings under the irradiation conditions examined. the results obtained in this study are as follows; (1)The rupture temperature of the irradiated PNC316 fuel cladding of MFA-1 was as same as that of our previous works for the fluence range up to 2.13
10
n/m
. There was no noticeable decrease in rupture temperature with increasing fluence in lower hoop stress region(
100MPa). (2)The rupture temperature of the irradiated 15Cr-20Ni fuel cladding of MFA-2 was almost as same as that of as-received cladding for the hoop stress range up to about 200MPa. The rupture temperature did not decrease significantly with fluence. (3)The rupture temperature of the irradiated PNC316 cladding tested at hoop stress 69MPa, which was the design hoop stress for MONJU fuel, was 1055.6
C. This suggested that the design cladding maximum temperature limit for MONJU (830C) was conservative. (4)There was no obvious relation between rupture temperature, swelling and microstructural change during transient heating under the irradiation conditions examined.
*
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).
; ;
JNC TN9400 2000-041, 29 Pages, 2000/03
JNC-TN9400-2000-041.pdf:1.18MB
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.
*
JNC TN9440 2000-002, 157 Pages, 2000/02
JNC-TN9440-2000-002.pdf:5.44MB
This report summarizes the operating and irradiation data of the experimental reactor "JOYO" 33rd cycle, and estimates the 34th cycle irradiation condition. Irradiation tests in the 33rd cycle are as follows: (1)B-type irradiation rig (B9) (a)High burn up performance tests of "MONJU" fuel pins, advanced austenitic steel cladding fuel pins, large diameter fuel pins, ferrite steel cladding fuel pins and large diameter annular pellet fuel pins (b)Mixed carbide and nitride fuel pins irradiation tests (in collaboration with JAERI) (2)C-type irradiation rig (C4F) (a)High burn up performance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (3)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (4)Absorber Materials Irradiation Rig (AMIR-6) (a)Run to absorber pin's cladding breach (5)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (6)Core Materials Irradiation Rig (CMIR-5-1) (a)Core materials irradiation tests (7)Structure Materials Irradiation Rigs(SMIR) (a)Material irradiation tests (in collaboration with universities) (b)Surveillance back up tests for "MONJU" (8)Upper core structure Irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect. The maximum burnup driver assembly "PFD516" reached 64,300MWd/t (pin average).
; ; Mizuta, Shunji
JNC TN9400 2000-023, 126 Pages, 2000/02
JNC-TN9400-2000-023.pdf:2.94MB
Modified 316 and 15Cr-20Ni base austenitic stainless steels had been developed by Japan Nuclear Cycle Development lnstitute as the candidate materials for Monju and Demonstration fast breeder reactor. Previously, irradiation creep correlation of modified 316 and 15Cr-20Ni had been evaluated using pressurized tubes irradiated in FFTF/MOTA. 0n the other hand, for other austenitic S.S. developed abroad, it was reported that irradiation creep behavior of fuel pin could not be sufficiently described using results of pressurized tube experiments. ln this study, irradiation creep properties of modified 316 and 15Cr-20Ni fuel pins (MFA-I, 2) irradiated in FFTF were evaluated. And irradiation deformation of MFA-1, 2 fuel pins were estimated using the irradiation creep correlation based on MOTA data. The results are summarized as follows : (1)Irradiation creep compliance B
calculated from MFA-I, 2 data are 5.6 15.010
[(I0
n/m, E>0.1Mev)
(MPa)], Which are larger than B based on MOTA data of 2.26.410 and are within the range of B of other austenitic S.S. abroad. (2)Creep-swelling coupling coefficient D derived from MFA-1, 2 data tend to decrease with increasing swelling rate. And the range of D based on MFA-1, 2 data include values calculated from MOTA data of 3.88.210
[(MPa)] and for other austenitic S.S. abroad. (3)As the result that irradiation creep deformation of MFA-1, 2 fuel pins could be appropriately estimated using the irradiation creep correlation derived from MOTA data, it is considered that the creep, correlation based on MOTA data can be applied to estimation of fuel pin deformation.
Mizuta, Shunji; ;
JNC TN9400 99-082, 60 Pages, 1999/10
The density measurement of the internal creep specimens irradiated in FFTF/MOTA (Fast Flux Test Facility / Material open Test Assembly) was conducted MMF (Materia1 Monitoring Facility) and accurate separation of swelling strain from total strain leaded in the derivation of the irradiation creep coefficients. Irradiation creep coefficients for PNC 316, 15Cr-20Ni base S.S. and 14Cr-25Ni base S.S. were systematically expressed, while thermal creep coefficients K, under irradiation were separately expressed for above three steels. The results obtained are follows, (1)The effect of stress induced swelling was recognized in the temperature range from 405 to 605C. The swelling in high stress specimens have a tendency to increasing swelling. (2)The irradiation creep coefficients derived from PNC316 and l5Cr-20Ni are similar to that of derived from 20%CW316S.S., CW316Ti and CW15-15Ti which were reported by other authors. (3)The irradiation creep coefficient derived from gas pressurized tube irradiation using FFTF/MOTA expressed appropriately irradiation creep strain from fuel pins using FFTF/MFA-2(15Cr-2ONi base S.S.).
Bottcher, J. T.
PNC TN9440 97-011, 215 Pages, 1997/06
J.H.Bottcher started his intemational Fellow position at PNC on March 25, 1996.During his 15 months in PNC he worked in the Irradiation Section of the Experimental Reactor Division. There he worked on conceptual design reviews and related the US irradiations methodology to the members. His work extended to other Divisions at OEC and Tokai Works, mainly related to fuel development and irradiation performance. In these efforts he published two papers, wrote a desip review document, and presented six lectures on irradiated fuels and materials. In addition he participated in coordinating a new four year PNC/DOE collaborative program on irradiated steels characterization.
Hunter
PNC TN9410 97-057, 106 Pages, 1997/05
This study was based on a 'pancake' type fast reactor core design of 600 MW(e), which had been optimized for Pu burning with a feed Pu vector appropriate to once-through irradiation of MOX fuel in a PWR. The purpose of the study was to investigate the effects of varying the Pu vector, examining various methods of offsetting the effects of such a change, and finally to produce fuel cycles optimized for the different qualities of Pu vector within the same basic design. In addition to the reference (once-through) Pu vector, two extreme Pu vectors were examined: high quality Pu from military stockpiles; low quality Pu corresponding to the equilibrium point of multiple recycling in a Pu burning fast reactor. Variations in Pu quality were overcome by changing the fuel inventory - replacing some of the fuel by diluent material, and altering the fuel pin size. Using absorber material (
B
C) as diluent improves the rod worth shutdown margin but degrades the Na void and Doppler safety parameters, a non-absorber diluent has the opposite effects, so a mix of the 2 material types was used to optimize the core characteristics. Of the non-absorber diluent materials examined, ZrH gave significantly better performance than all others; 
BC was the second choice for non-absorber diluent, because of its compatibility with BC absorber. It was not possible to accommodate the lower quality (multi-recycled) Pu vector without a significant increase in the fuel pin volume. It was not generally possible, especially with the increased fuel pin size, to achieve positive rod worth shutdown margins - this was overcome by increasing the number of control rods. For the higher quality Pu vectors to maintain ratings within limits, it was necessary to adopt hollow fuel pellets, or else to use the diluent material as an inert matrix in the fuel pellets. It proved possible to accommodate both extremes of Pu vector within a single basic design, maintaining ...
Arai, Yasuo; Iwai, Takashi; ; Okamoto, Yoshihiro; Shiozawa, Kenichi;
JAERI-Research 96-009, 17 Pages, 1996/02
JAERI-Research-96-009.pdf:1.09MB
no abstracts in English
Power Reactor and Nuclear Fuel Development Corporation
PNC TN9360 95-002, 98 Pages, 1995/11
no abstracts in English
Power Reactor and Nuclear Fuel Development Corporation
PNC TN9360 95-001, 104 Pages, 1995/11
no abstracts in English
Okada, Toshio; *; *; *; *; Miyake, Osamu; Sasaki, Shuichi
PNC TN1410 94-006, 57 Pages, 1994/11
no abstracts in English
Power Reactor and Nuclear Fuel Development Corporation
PNC TN9360 94-002, 100 Pages, 1994/02
no abstracts in English
Power Reactor and Nuclear Fuel Development Corporation
PNC TN9360 94-001, 95 Pages, 1994/02
no abstracts in English
Isozaki, Kazunori; ; Ito, Hideaki; ; Chatani, Keiji; ;
PNC TN9520 93-006, 198 Pages, 1992/11
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