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JAEA Reports

Irradiation tests report of the 33rd cycle in "JOYO"

*

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).

JAEA Reports

RB2 Pre-test Calculation using PAPAS-2S based on a Preliminary Post-test Calculation of the RB1 Test

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PNC-TN9410 98-058, 12 Pages, 1998/06

PNC-TN9410-98-058.pdf:0.84MB

Based on the RB1 test result in the CABRI-RAFT Program, it was agreed between the partners to perform the RB2 test which aims at observation of molten fuel ejection into the coolant channel at further fuel melting and at confirmation of coolability of ejected fuel. In this study, a preliminary post-test calculation for the RB1 test is performed first to reflect the fuel thermal condition expected for the pins with the special artificial defect preparation. Pre-test calculations for the RB2 test are then performed based on the results of this RB1 calculation. Power and coolant flow histories as well as the axial location of defect were selected as parameters in this study and a set of test condition is proposed which is believed to be most suitable to fulfill the test objectives.

JAEA Reports

Fuel pin failure threshold under the slow TOP condition; Survey on the existing In-pile tests and investigation of the FCMI mitigation mechanism

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PNC-TN9410 98-057, 55 Pages, 1998/05

PNC-TN9410-98-057.pdf:3.99MB

Existing data of in-pile ramp-type transient-overpower tests (slow TOPs hereafter), such as those of the CABRI-2 and CABRI-FAST tests, the EBR-II TOPI-1E test and the former TREAT tests, were extensively surveyed and this led to a global interpretation which provided a consistency among the tests. Through this study, a basic fuel pin failure mechanism was comprehended and it was confirmed that fuel pins with low to intermediate smear density have a very high failure threshold with significant mitigation effects against fuel-cladding mechanical interactions. Such high failure threshold of low to intermediate smear density fuel is considered to be attributed to the following three effects: (1)absorption of fuel thermal expansion and fuel swelling by void space (porosity or cracks) within the fuel, (2)mitigation of fuel swelling by an early gas escape into the free volume, and (3)mitigation of molten cavity pressurization upon fuel melting. These effects were refrected to the analytical model of the transient fuel behavior code PAPAS-2S. Application of this improved PAPAS-2S model to representative slow TOP tests provided results consistent with the test data, suggesting that the above-mentioned consideration is valid.

JAEA Reports

Development of the flow control irradiation facility for JOYO

Soroi, Masatoshi;

PNC-TN9410 98-050, 57 Pages, 1998/05

PNC-TN9410-98-050.pdf:1.58MB

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$$ell$$/min flow rate and 265kPa pressure in 550$$^{circ}$$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 ...

JAEA Reports

A Study of iodine diffusion in rare gases(III)

Sagawa, Norihiko*

PNC-TJ9613 97-002, 95 Pages, 1997/10

PNC-TJ9613-97-002.pdf:2.22MB

The diffusion coefficient of cesium iodide vapor in rare gases was determined by a modified Stefan's method. The rare gas in a diffusion column was saturated with vapor of the cesium iodide, crystals of which were heated to melt at the bottom of the column. By opening a valve united at a top of the column, the vapor diffusing through the column was transferred with the carrier rare gas to an ionization sensor. The concentration of cesium iodide in the carrier gas was continuously monitored with the sensor. The diffusion coefficient was determined by analyzing the transient response of the concentration. Increasing tendency with temperature is observed in the coefficients obtained in argon, kripton and xenon at temperatures between 631 and 691 $$^{circ}$$C and no significant difference among the coefficients in argon, krypton and xenon.

JAEA Reports

A Study of iodine diffusion in rare gases(II)

Sagawa, Norihiko*

PNC-TJ9613 97-001, 90 Pages, 1997/10

PNC-TJ9613-97-001.pdf:1.69MB

An ionization sensor. which ionizes iodine vapor on a heated filament and collects ionized iodine on a collector at positive potential, was made on an experimental basis. Iodine vapor in rare gas was determined with using the sensor on the line and the characteristic of the sensor was examined. Iodine vapor was generated from iodine crystals in an iodine evaporator and transferred to the sensor with carrier-rare gas. The iodine vapor was continuously monitored by the sensor and trapped in solution of sodium hydroxide. The amount of iodine in the solution was determined by chemical analyses. The integrated value of ion current was compared with the collected amount of the iodine. A proportional relation is observed between the collected amount and the integrated value obtained from the sensor with a platinum collector, but not found between the amount and the value obtained from the sensor with a stainless steel collector.

JAEA Reports

None

;

PNC-TN8410 97-066, 304 Pages, 1997/02

PNC-TN8410-97-066.pdf:121.03MB

None

JAEA Reports

None

Kato, Masato

PNC-TN8600 94-005, 132 Pages, 1994/08

PNC-TN8600-94-005.pdf:7.95MB

no abstracts in English

JAEA Reports

None

Kato, Masato

PNC-TN8600 94-004, 184 Pages, 1994/08

PNC-TN8600-94-004.pdf:9.48MB

no abstracts in English

JAEA Reports

None

PNC-TN1410 94-052, 181 Pages, 1994/06

PNC-TN1410-94-052.pdf:5.58MB

no abstracts in English

JAEA Reports

Planning study of in-pile loop tests for the evaluation of fission product transport

Nakagiri, Toshio; ; Ohno, Shuji; ; *; Koyama, Shinichi;

PNC-TN9510 94-001, 246 Pages, 1994/05

PNC-TN9510-94-001.pdf:6.38MB

None

JAEA Reports

Development of in-vessel source term analysis code, TRACER

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PNC-TN9410 93-015, 163 Pages, 1993/01

PNC-TN9410-93-015.pdf:4.82MB

To evaluate the species and quantities of fission products (FPs) released into the cover gas in an LMFBR during the fuel pin failure accidents (In-vessel Source Term), the computer code TRACER (Transport phenomena of Radionuclides for Accident Consequence Evaluation of Reactor) has been developed. The code can analyze mechanistically the physical and chemical phenomena of FPs release and transport behavior from the failed fuel to the cover gas through sodium coolant. The code validation was performed through the sample calculations for the results of Mo17C/6 in-pile source term experiment. In the beginning of this study, the production and transport phenomena of bubbles of the noble gas and volatile FPs in the coolant, and mass transfer between the bubble and coolant were investigated, and the functions needed for the code were clarified. After that, several mechanistic models for the bubble behavior were developed, and the programs were coded. From the results of the sample calculations, it was confirmed that the code was capable of simulating the FP transport phenomena in the primary system.

JAEA Reports

Introduction of Nuclear Instrumentations and Radiation Measurements in Experimental Fast Reactor 「JOYO」

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PNC-TN9420 92-005, 83 Pages, 1992/04

PNC-TN9420-92-005.pdf:2.17MB

This report introduces the nuclear instrumentation system and major R&D (research and development) activities using radiation measurement techniques in Experimental Fast Reactor "JOYO". In the introduction of the nuclear instrumentation system, following items are described; (1)system function (2)roles as a reactor plant equipment (3)specifications and charactelistics of neutron detectors, (4)construction and layout of the system. For reactor dosimetry at various irradiation tests and surveillance tests, multi-foil method employed in "JOYO", neutron fluence evaluation using activation foils and HAFM (Helium Accumulation Fluence Monitor) under development are described briefly. The failed fuel detection system and some experimental equipments using radiation measurement techniques are also introduced here with main results obtained by a series of fuel failure simulation experiments. In addition, following R&Ds are picked up as some examples based on radiation measurement technology; (1)burn-up measurement of spent fuel subassembly (2)measurement and evaluation of radiation source distributions (radioactive corrosion products)

JAEA Reports

None

PNC-TN1410 92-026, 113 Pages, 1992/01

PNC-TN1410-92-026.pdf:11.01MB

no abstracts in English

JAEA Reports

Analysis of hypothetical core disruptive accident in prototype fast breeder reactor Monju (I); Analysis of HCDA initiating phase by SAS3D code

*; *; Aoi, Sadanori*

PNC-TN941 82-74VOL1, 151 Pages, 1982/03

PNC-TN941-82-74VOL1.pdf:7.53MB

A study of hypothetical core disruptive accidents (HCDAs) in the prototype fast breeder reactor Monju (714 MWt) has been conducted by using the SAS3D$$^{#}$$ accident analysis code. A loss-of-flow (LOF) due to the loss of off-site power and a transient overpower (TOP) due to control assembly withdrawal, both at rated power, are considered as the HCDA initiators with a postulated total failure of the reactor shutdown system. The accident scenarios of each postulated anticipated transient without scram are studied for the three burnup stages of Monju: the beginning-of-initial cycle (BOIC) ; a beginning-of-equilibrium cycle (BOEC); and an end-of-equilibrium cycle (EOEC). The neutronics data used in this study has been obtained by a 3-dimensional HEX-Z diffusion code and the first order perturbation calculations. The reactivity coefficients used in this study are the design nominal values without taking into account their uncertainties. The nominal design value of the maximum positive sodium void worth in Monju is a relatively small value of 2.5$ in the EOEC core. In the 2 cents/sec TOP, the reactor power shows a sudden increase following the onset of FCIs (Molten-Fuel/Coolant Interactions) in high-powered fuel assemblies but the maximum power level reached is less than 5 times the rated power and due to the fuel sweepout negative reactivity in the FCI fuel assemblies, the reactor is shutdown within 0.1 sec at the latest after the first FCI onset. The extent of damaged fuel assemblies is largest in the clean (FP-gas free) BOIC core in which the radial power peaking is smaller than in BOEC and EOEC cores, and about 17% of the fuel assemblies are damaged in the central region of the core. In the equilibrium cycle cores the damage extents are limited to about 5% core-center assemblies and this is smaller than in the BOIC core because of the larger radial power peaking and the rapid fuel sweepout reactivity insertion accelerated by the FP-gas pressure in the ...

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