Fuel Processing Facility Research Articles

Novel thermometry approaches to facilitate safe and effective monitoring of nuclear material containers

Many established nuclear power producing countries are currently decommissioning first and increasingly second-generation power producing plants and fuel processing facilities. This has led to a growing inventory of different containers and packages containing radioactive waste and other nuclear materials, as well as storage of spent fuel. Here we describe research to establish in-situ yet remote health monitoring techniques based on novel temperature measurement methods for different containers and racks used to hold different nuclear waste forms, special nuclear materials and spent fuel.

Development of prototype induced-fission-based Pu accountancy instrument for safeguards applications

Prototype safeguards instrument for nuclear material accountancy (NMA) of uranium/transuranic (U/TRU) products that could be produced in a future advanced PWR fuel processing facility has been developed and characterized. This is a new, hybrid neutron measurement system based on fast neutron energy multiplication (FNEM) and passive neutron albedo reactivity (PNAR) methods. The FNEM method is sensitive to the induced fission rate by fast neutrons, while the PNAR method is sensitive to the induced fission rate by thermal neutrons in the sample to be measured. The induced fission rate is proportional to the total amount of fissile material, especially plutonium (Pu), in the U/TRU product; hence, the Pu amount can be calibrated as a function of the induced fission rate, which can be measured using either the FNEM or PNAR method. In the present study, the prototype system was built using six 3He tubes, and its performance was evaluated for various detector parameters including high-voltage (HV) plateau, efficiency profiles, dead time, and stability. The system's capability to measure the difference in the average neutron energy for the FNEM signature also was evaluated, using AmLi, PuBe, 252Cf, as well as four Pu-oxide sources each with a different impurity (Al, F, Mg, and B) and producing (α,n) neutrons with different average energies. Future work will measure the hybrid signature (i.e., FNEM×PNAR) for a Pu source with an external interrogating neutron source after enlarging the cavity size of the prototype system to accommodate a large-size Pu source (~600g Pu).

Investigation of n-pentane pyrolysis at elevated temperatures and pressures in a variable pressure flow reactor

Thermally cracking fuel in a fuel processing facility or undergoing pyrolysis in the anode channel of a fuel cell can lead to coke deposition. The formation of coke occurs via chemical kinetic pathways that are largely dependent on poly aromatic hydrocarbons (PAHs). The formation of PAH⿿s largely depends on species which include ethylene, acetylene, butadiene, benzene and toluene. Pentane is a relatively simple structure for modeling purposes, yet large enough that it produces some of the relevant chemistry of real fuels. Experiments were conducted in a newly designed variable pressure flow reactor (VPFR) built to process both liquid and gaseous reactants at pressures to 4.0MPa and temperatures to 1373K. The VPFR was initially validated via ethane pyrolysis experiments and comparing the results with existing literature data. Pentane experiments were done over a range of temperatures (923⿿1073K), pressures (1⿿2.0MPa) and residence times (30ms⿿40s). The product selectivity appeared to be greatest influenced by pressure and level of fuel conversion. Pressure increases ethylene production but has an inverse effect on ethane production. Soot precursors appeared at higher levels of conversion. The results of the experiments were compared to an existing chemical kinetic mechanism with good agreement.

Transient response and radiation dose estimates for breaches to a spent fuel processing facility

This paper describes the analysis of the design basis accident for Idaho National Laboratory Fuel Conditioning Facility (FCF). The facility is used to process spent metallic nuclear fuel. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities.

Proliferation Resistance of a Hypothetical Sodium Fast Reactor under an Assumed Breakout Scenario

The Proliferation Resistance and Physical Protection (PR&PP) Working Group of the Generation IV International Forum conducted a high-level pathway analysis of a hypothetical sodium fast reactor and integral fuel processing facility (called collectively the Example Sodium Fast Reactor, or ESFR), as a test of the effectiveness of its analysis methodology. This paper presents the results of the analysis based on the breakout scenario. Four representative strategies were chosen for analysis: diversion of low-enriched uranium feed material, two different types of misuse of the reactor facility, and misuse of the fuel processing facility. A high-level pathways analysis was conducted for each strategy to determine relative ranking of the proliferation-time measure, specifically as it applies to the postbreakout period.

D132 燃料加工施設ケーブルの熱劣化試験(OS-11:火災安全(III))

D132 燃料加工施設ケーブルの熱劣化試験(OS-11:火災安全(III))

Uranium (UO22+) Retention Property of Degraded n-dodecane by Acidic Radiolysis in the Purex Process

Radiation effects on the uranium retention property of degraded n-dodecane by acidic radiolysis in the Purex process were studied. The absorbed dose of n-dodecane varied from 24.04 to 2403.69 kGy (5 to 500 Wh/dm3) . When the absorbed dose exceeded 48.07 kGy (10 Wh/dm3), the uranium (UO22+) was retained in degraded n-dodecane even without TBP as extractant. An empirical correlation of the uranium retention was also derived. This correlation is a very effective one because it can especially be used for the prediction of the amount of retained uranium in the degraded n-dodecane. Data obtained through this work should be useful for the process design under solvent degradation conditions in nuclear fuel processing facilities especially in reused solvent.

Validation of Simplified Evaluation Models for First Peak Power, Energy, and Total Fissions of a Criticality Accident in a Nuclear Fuel Processing Facility by TRACY Experiments

Simplified evaluation models are developed at the Japan Atomic Energy Research Institute (JAERI) to predict the first peak power, energy, and total fission numbers during a criticality accident for design and installation of a criticality alarm system and for quick response with measures to avoid excessive exposure of the general public. These models were first derived in previous papers only from theoretical considerations employing one-point reactor kinetic neutron behavior and thus are applicable to any geometrical shape of vessel containing fissile solution. Applicability concerning nuclide composition comes essentially from using empirical equations describing specific heat and density to give simplified forms of the models. The models developed originally for a stepwise reactivity insertion mode are shown in the current paper to approximately stand for the ramp reactivity insertion mode by giving their theoretical formation and are validated by applying experimental data from JAERI’s Transient Experiment Critical Facility (TRACY) on a low-235U-enriched uranium nitrate solution as well as CRAC experiments on high-235U-enriched uranium nitrate solution together with past accident data, including the most recent JCO accident.

｀２４´Ｎａの体内放射能測定とモニタリングデータを用いたＪＣＯ臨界事故における従業員等の被ばく線量評価

A criticality accident occurred at nuclear fuel processing facility of JCO Company Ltd. at 10:35 local time on September 30, 1999. In this accident, the doses to few workers were evaluated directly with individual dosimeters. Japan Nuclear Cycle Development Institute (JNC) estimated the dose to the JCO workers in cooperation with the Criticality Accident Investigation Committee. JNC carried out the dose evaluation using the following methods.(1) Dose evaluation based on 24Na activities produced in the human body due to neutron capture reaction, which were measured with whole body counter.(2) Dose evaluation by the calculation from the change of the radiation dose rate which is estimated by the area monitors, and the behavior of the employee at the criticality accident.To validate the results of the dose evaluation, dose calculation was carried out to several workers whose dose had been measured with individual dosimeters or a whole body counter. As the result, the dose calculation technique was almost applicable for the dose estimation of JCO workers.

海外の核燃料施設における臨界事故の原因分析 ＪＣＯウラン加工場臨界事故との類似性

On September 30, 1999, a criticality accident occurred at the JCO's uranium processing plant, which resulted in the first nuclear accident involving a fatality, in Japan, and forced the residents in the vicinity of the site to be evacuated and be sheltered indoors. Before the JCO accident, 21 criticality accidents have been reported at nuclear fuel processing facilities in foreign countries. The present paper describes the overall trends observed in the 21 accidents and discusses the sequences and causes of the accidents analyzed in terms of similarities to the JCO accident. Almost all of them occurred with the uranium or plutonium solution and in vessels/tanks with unfavorable geometry. In some cases, the problems similar to those observed in the JCO accident were identified: violations of procedures and/or technical specifications for improving work efficiencies, procedural changes without any application to and permission from the regulatory body, lack of understanding of criticality hazards, and complacency that a criticality accident would not occur.

Applications of dendrochronology and sediment geochronology to establish reference episodes for evaluations of environmental radioactivity

Dendrochronology and sediment geochronology have been used to demonstrate retrospective monitoring of environmental radioactivity at United States Department of Energy (DOE) sites. 14C in annual growth rings of sagebrush preserved the temporal and spatial patterns of 14C resulting from dispersion downwind of a nuclear fuel processing facility at the Hanford Site in Washington State. As far as 10 km downwind of the facility, 14C concentrations were significantly higher in growth rings formed during a fuel processing episode than in rings produced during preoperational or postoperational episodes. An episode of uranium mill tailings deposition in pond sediments at the Grand Junction Office in Colorado was reconstructed using 210Pb geochronology constrained by a marker of peak 137Cs fallout. Uranium concentrations in ponds sediments deposited after the processing episode provide a reasonable cleanup standard. These reference episodes of environmental radioactivity reconstructed from measurements taken within contaminated environments can improve or replace reference area data as baseline information for dose reconstructions, risk assessments, and the establishment of cleanup standards.

Shield Verification Testing at the Idaho Chemical Processing Plant — Fuel Processing Facility

Abstract A Shield Testing Program is being conducted at a new nuclear facility to verify that shield walls, doors, and windows meet the design criteria for radiation attenuation. A modified Teletherapy Unit is being utilized as a test unit. This unit has been successfully used for shield testing during the construction of two other major facilities. A collimated beam from a Co60 source is exposed on one side of the shield being tested, and the attenuated radiation is measured with a standard radiation detector. Detailed documentation is completed for each area tested. Detailed procedures are required for operation of the test unit. No significant personnel exposure has been recorded during any of the shield testing. The Co60 test method is a known, proven technique that provides the required penetration and the capability to cover relatively large areas with a single exposure.

Mutagenic activity of surface waters adjacent to a nuclear fuel processing facility.

Surface waters adjacent to a nuclear fuel processing facility were extracted, using XAD-resin adsorption followed by solvent elution, and the extracts were assayed for mutagenic potential by the AmesSalmonella-mammalian microsome test. Dose-related mutagenic responses with TA102 (+ S9) were produced with the extracts of water samples obtained from a creek receiving waste-water from the processing facility (specific mutagenic activities of 7,250 to 8,250 net revertants per L equivalent of water). The creek water extracts were not mutagenic with TA102 in the absence of S9, or with any other tester strain (i.e., TA97, TA98, TA100, and TA1535) in the presence or absence of S9. Surface water samples downstream and upstream of this creek were not mutagenic; apparently indicating the lack of persistence of the observed mutagenicity. The major constituent in the mutagenic creek water extracts was identified as tributylphosphate (TBP) by gas chromatography-mass spectrometry. However, TBP was not mutagenic with TA102 (+ S9) at doses ranging from 196 μg/plate to 9.8 ng/plate. Because tester strain TA102 detects oxidative mutagenesis due to x-rays and ultraviolet radiation, it is possible that the observed mutagenicity of creek water extracts was due to radionuclides complexed to TBP.

RADIOIODINE RELEASE INCIDENT AT THE SAVANNAH RIVER PLANT.

Abstract—A release of 153 c of radioiodine from a Purex fuel processing facility, most of which occurred May 30 through June 3, 1961, temporarily increased the level of I131 in the environment of the Savannah River Plant (SRP). Prevailing southwest winds and atmospheric inversions dispersed the radioiodine mainly to the northeast of the Plant. Theoretical dispersion equations were verified by measurements of I131 in air at distances up to 25 miles from the Plant. I131 on vegetation decayed with an apparent half-life of 5 days, as compared to the physical half-life of 8 days. This is attributed to dilution of the activity by new vegetative growth. The reduction of radioiodine in milk closely followed the apparent half-life in grass. Laboratory tests with iodine-contaminated vegetation showed that rainfall removed little or no iodine. Consumption of milk in the area of highest deposition could have resulted in a thyroid dose up to 1.2 rems for a child. Empirical relationships relating the transfer of radioiodine from air to grass and then to milk are compared with those determined earlier at SRP and elsewhere.