后处理设施的红油爆炸故障树分析

目前世界上几乎所有的乏燃料后处理设施均选用Purex流程。红油爆炸事故是Purex流程后处理设施安全研究的焦点问题之一。本文对典型后处理设施共去污分离循环工段的中放废液蒸发器的红油爆炸事故进行了事件树 故障树分析。结果显示,红油爆炸事故的发生频率极低;人因失误和共因失效对整个红油爆炸事故的贡献很大。应采取相应措施,以预防事故的发生。     

Co-processing of uranium and plutonium for sodium-cooled fast reactor fuel reprocessing by acid split method for plutonium partitioning without reductant

A solvent extraction flowsheet for Pu partitioning, based on the acid split method without reductant, originally proposed by the Oak Ridge National Laboratory (ORNL), was tested for sodium-cooled fast reactor fuel reprocessing. To enhance resistance to nuclear proliferation, a flowsheet for co-processing was developed that controls Pu content in the products while avoiding Pu polymerization and formation of a third phase during extraction. In this method, Pu is partitioned using the difference in distribution coefficients of U and Pu. It is effective for selective Pu stripping from U at low temperatures and HNO₃ concentrations. The flowsheet with a supply of 0.15 mol/dm³ HNO₃ solution at 21°C for Pu partitioning was tested experimentally using miniature centrifugal contactors and a highly radioactive solution. Neither a Pu(IV) polymer nor a third phase was observed during the experiment. The Pu content in the U/Pu product increased to 2.28 times that in the feed solution. The leakage ratio of Pu to the U product was slightly less in the U stripping section. Some fission products (FPs) were effectively decontaminated; e.g., decontamination factors (DFs) of Cs in U/Pu and U products were 4.51×10⁵ and 2.42×10⁵, respectively.     

Application of “Hydration Model” to Evaluate Gas Phase Transfer of Ruthenium and Technetium from Reprocessing Solutions

In order to evaluate the amounts of gas phase transferred ruthenium (Ru), and technetium (Tc), simulations were made for the continuous evaporator used in a reprocessing plant to concentrate high level liquid waste. The concentrations and activities of nitric acid and water, which controlled the reaction rate and gas-liquid equilibrium in the evaporator solution, were evaluated using the previously developed “Hydration Model”. When the feed solution contained 2.7 M (=mol/dm³) of nitric acid, the nitric acid concentration in the evaporator solution reached its maximum at the concentration factor (CF) of 6 (CF: concentration ratio of FPs in evaporator and feed solutions). The activities of nitric acid and water were saturated at values of 0.01 and 0.43, respectively, after the CF reached 6. The simulation predicted decontamination factors DFs of 2×10⁵ and 8×10³ for Ru and Tc, respectively, for a typical evaporation conditions with an operational pressure of 6,700 Pa, and FPs of 0.02 to 1.4 M. The simulation results agreed with the verification experiment, which will be described in next paper, within a factor of 2 for the amount of gas-phase transferred Ru during evaporation. The factor for the amount of gas-phase transferred Tc was estimated as 5 from the measurement error in the gas-liquid equilibrium constant.     

Isotopic composition of plutonium in the bottom deposits of the Norwegian Sea and Greenland Sea and identification of the sources of contaminationhearder

The results of measurements of the isotopic atomic ratios²³⁸Pu/²³⁹Pu,²⁴⁰Pu/²³⁹Pu,²⁴¹Pu/²³⁹Pu,²⁴²Pu/²³⁹Pu in samples of bottom deposits are presented. The role of the main sources of plutonium contamination of the region is examined. It is shown that the contribution of the α activity of industrial plutonium (^238–240Pu) in bottom sediments from the European reprocessing plants, mainly, Sellafield (Great Britain), on the background due to global fallouts is 20 to 55% of the total α activity and 70–95% for the β activity (²⁴¹Pu). No escape of reactor and weapons plutonium from the nuclear powered submarine “Komsomolets” into the environment was observed. The lower limit of sensitivity of the method is 10⁻¹⁴g of plutonium in a sample, which corresponds to a plutonium α activity of 0.02 mBq in a 10 g sample of bottom deposits, 4 tables, 26 references. Translated from Atomnaya énergiya, Vol. 87, No. 4, pp. 286–294. October, 1999.     

Nuclear Analysis of PIE Data of Irradiated BWR 8×8-2 and 8×8-4 UO2 Fuel Assemblies

The measured pellet average inventories of actinides and fission product nuclides on the fifteen samples taken from a three-cycle irradiation BWR 8×8-2 UO₂ assembly were compared with those of assembly burnup calculations using a collision probability method (SRAC) with the JENDL-3.2 nuclear data library. The present calculations overestimate the inventories of ²³⁵U, well reproduce those of ²³⁹Pu and ²⁴⁰Pu, yet underestimate those of ²³⁶U, ²³⁷Nd, ²³⁸Pu, ²⁴¹Pu, and ²⁴²Pu. The inventories of minor actinides are underestimated by the present analysis except for ²⁴¹Am. The major FP nuclides contributing to neutron absorption such as Nd, Cs, Eu, and Sm are almost well reproduced by the present calculations. The measured pellet average burnups and major actinide inventories on the twenty samples taken from four BWR 8×8-4 UO₂ assemblies were also compared with those of the burnup calculations using SRAC and a continuous energy Monte Carlo burnup analysis code (MVP-BURN). Most of the calculated pellet average burnups of both codes agree with the measurements within the range of ±10%. The general trends of the measured pellet radial distributions of actinide and FP nuclides on six samples of the 8×8-4 UO₂ assemblies were well reproduced by the burnup calculations of MVP-BURN.     

Diffusion Coefficient of Fission Gas in Uranium Dioxide Powder Formed by Carbon Dioxide Oxidation of Uranium

In connection with a program to study the behavior of punctured fuel elements for the Tokai Atomic Power Reactor, the diffusion coefficient of fission gas in uranium oxide powder formed by CO₂ oxidation of U was determined by post-irradiation experiment, in which the fractional release of fission gas during isothermal heating of the powder was measured. The U was oxidized at 600° and 700°C, and in both cases the O/U ratio of the oxides, measured gravimetrically, was 2.0. The diffussion coefficients in the oxide powder formed by oxidation at 600°C were found to be 1.4× 10-²⁰, 1.3×10-¹⁹, 1.1×10¹⁸ and 1.0×10-¹⁷, cm²sec-¹, respectively at 450°, 550°, 650° and 750°C, and in the oxide powder formed at 700°C, 7.4×10-¹⁹ and 3.6×10-¹⁶cm²sec-¹ at 600° and 700°C, respectively. Activation energies calculated for the two oxide powders were comparatively low.     

Separation of Antimony-125 in Fission Products

Most of the known radioactive nuclides of antimony produced by neutron irradiation of uranium have fission yields below 1% and have half-lives below 60 days. An exception is ¹²⁵Sb with a half-life of 2.7 yr, which raise its relative importance among the fission products with lapse of time after irradiation, and after 1 yr of cooling, its radioactivity is no longer negligible. This circumstance has led to its being separated from such sources as fall-out. No studies have so far been reported on using the nitrate system for this separation, though it is utilized in the reprocessing of spent fuel and in the dissolution of uranium samples. The present work describes a method of separating ¹²⁵Sb from fission products with use made of silica gel—nitric acid system, and an example of its application to the separation of ¹²⁵Sb from the spent fuel of JPDR-1. The fuel was irradiated from Oct. 1963 to Sep. 1969. The amount of ¹²⁵Sb measured after separation was (1.7± O.19)×10^?1Ci/gU at June 1972.     

Reliability analysis of shutdown system

This paper presents the results of reliability analysis of Shutdown System (SDS) of Indian Prototype Fast Breeder Reactor. Reliability analysis carried out using Fault Tree Analysis predicts a value of 3.5 × 10⁻⁸/de for failure of shutdown function in case of global faults and 4.4 × 10⁻⁸/de for local faults. Based on 20 de/y, the frequency of shutdown function failure is 0.7 × 10⁻⁶/ry, which meets the reliability target, set by the Indian Atomic Energy Regulatory Board. The reliability is limited by Common Cause Failure (CCF) of actuation part of SDS and to a lesser extent CCF of electronic components. The failure frequency of individual systems is <1 × 10⁻³/ry, which also meets the safety criteria. Uncertainty analysis indicates a maximum error factor of 5 for the top event unavailability.     

Techniques de mesure des coefficients d'heterodiffusion dans le plutonium. Application a la diffusion du cobalt dans le plutonium ϵ(cc)

The diffusion of ⁶⁰Co in bec ? plutonium has been studied by the sectioning method, and the following results have been obtained: D = 1.4 × 10^?3 exp (?9900/RT) cm²/sec over the temperature range 484–621°C. Cobalt diffuses rapidly in ? Pu. Since the diffusion coefficient does not change across the phase transition ? Pu (fcc) → ? (bcc), the diffusion mechanism must be dissociative in the two phases.     

Photochemically-Induced Valency Adjustment of Plutonium and Neptunium in Nitric Acid Solution Using Mercury Lamp

A photochemically-induced valency adjustment method has been studied to remove Np from the mixed nitric acid solutions of Pu and Np in connection with the Purex reprocessing. The valencies of Pu and Np ions were adjusted to be Pu(HI) and Np(V) under the initial conditions and their concentrations were 1x10^?4 and 1x10^?3 mol·dm^?3, respectively. The experiments were carried out under the various conditions changing the irradiation intensities of the Hg lamp in the various concentrations of HNO₃. It was found that the rates of the redox reactions of the Pu ions were significantly affected by the irradiated light as well as the acid strength. Under the irradiation of the 0.015 W Hg lamp in 3 M HNO₃ solution containing a tenfold excess of a hydroxylamine and hydrazine, more than 95% Pu(ID) was oxidized rapidly to Pu(IV) within 10 min irradiation and it remained at the same valency even after the continuous further irradiation.

On the other hand, the irradiation did not change the valency of Np(V) under the conditions studied. These valency conditions, i.e. Pu(IV) and Np(V), are appropriate for separating Np from Pu by the solvent extraction with TBP-n-dodecane.

The present results lead to the conclusion that the photochemical method has a high potential for removing Np from the mixed solution of Pu and Np. The photochemical redox reaction mechanisms of Pu and Np in the nitric acid solution were discussed from the stand-points of the thermodynamic and kinetic considerations related to the variation in their standard electrode potentials of the photo-excited ion species by the light irradiation.     

On the electrochemical formation of Pu–Al alloys in molten LiCl–KCl

Properties of Pu–Al alloys were investigated in connection with development of pyrochemical methods for reprocessing of spent nuclear fuel. Electroseparation techniques in molten LiCl–KCl are being developed in ITU to group-selectively recover actinides from the mixture with fission products. In the process, actinides are electrochemically reduced on solid aluminium cathodes, forming solid actinide–aluminium alloys. This article is focused on electro-chemical characterisation of Pu–Al alloys in molten LiCl–KCl, on electrodeposition of Pu on solid Al electrodes and on determination of chemical composition and structure of the formed alloys. Cyclic voltammetry and chronopotentiometry were used to study Pu–Al alloys in the temperature range 400–550 °C. Pu is reduced to metal in one reduction step Pu³⁺/Pu⁰ on an inert W electrode. On a reactive Al electrode, the reduction of Pu³⁺ to Pu⁰ occurs at a more positive potential due to formation of Pu–Al alloys. The open circuit potential technique was used to identify the alloys formed. Stable deposits were obtained by potentiostatic electrolyses of LiCl–KCl–PuCl₃ melts on Al plates. XRD and SEM–EDX analyses were used to characterise the alloys, which were composed mainly of PuAl₄ with some PuAl₃. In addition, the preparation of PuCl₃ containing salt by carbochlorination of PuO₂ is described.     

Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

An efficient dissolution process was established for future reprocessing in which mixed-oxide (MOX) fuels with high plutonium contents and dissolver solution with high heavy-metal (HM) concentrations (more than 500 g dm^?3) will be treated. This dissolution process involves short stroke shearing of fuels (～10 mm in length). The dissolution kinetics of irradiated MOX fuels and the effects of the Pu content, HM concentration, and fuel form on the dissolution rate were investigated. Irradiated fuel was found to dissolve as 10²–10³ times fast as non-irradiated fuel, but the rate decreased with increasing Pu content. Kinetic analysis based on the fragmentation model, which considers the penetration and diffusion of nitric acid through fuel matrices prior to chemical reaction, indicated that the dissolution rate of irradiated fuel was affected not only by the volume ratio of liquid to solid (L/S ratio) but also by the exposed surface area per unit mole of nitric acid (A/m ratio). The penetration rate of nitric acid is expected to be decreased at high HM concentrations by a reduction in the L/S ratio, but enhanced by shearing the fuel pieces with short strokes and thus enlarging the A/m ratio.     

Improvement of bitumen-waste product in leachability—Leachability of bitumen product containing BWR's evaporator concentrate

The method has been developed to improve the bitumen product which incorporates an evaporator concentrate from a BWR, with respect to the swelling and leaching. The leachability of the product has been measured by the method recommended by the IAEA. The swelling of the product is successfully prevented by the addition of calcium chloride. The specimen containing the waste up to Wa/B (Weight ratio of Na₂SO₄ + CaCl₂ and bitumen) = 60:40 shows no pronounced swelling, when it is immersed in water. The cumulative fractions of ¹³⁷Cs and ⁶⁰Co leached from a specimen which does not contain CaCl₂ are 0.65 and 0.2 at the leaching time of 30 days. On the hand, the corresponding value at 100 days for the specimen with calcium chloride addition is 5 × 10^?4 for ¹³⁷Cs and 1 × 10^?4 for ⁶⁰Co. The coating of the specimen surface with a fresh bitumen (5 mm thickness) reduces the leachability further. These results indicate that this method is effective to improve the bitumen product incorporating BWR's evaporator concentrate.     

Fundamental safety strategy against severe accidents on prototype sodium-cooled fast reactor

The accident categories of severe accidents (SAs) for prototype sodium-cooled fast reactor (SFR) which need proper measures were investigated through the internal event probabilistic risk assessment (PRA) and event tree analysis for the external event and six accident categories, unprotected loss of flow (ULOF), unprotected transient over power (UTOP), unprotected loss of heat sink (ULOHS), loss of reactor sodium level (LORL), protected loss of heat sink (PLOHS) and station blackout (SBO), were identified. Fundamental safety strategy against these accidents is studied and clearly stated considering the characteristics and existing accident measures of prototype SFR, and concrete measures based on this safety strategy are investigated and organized. The sufficiency of these SA measures is confirmed by comparing the evaluated core damage frequency (CDF) and containment failure frequency (CFF) to the target value, 1×10^?5 and 1×10^?6 per plant operating year, respectively, which were selected based on the IAEA's safety target. However, the target value of CDF and CFF should be satisfied considering all the SAs caused by both internal and external events. External event PRA for prototype SFR is now under evaluation and we set out to satisfy the target value of CDF and CFF considering both internal and external events.     

Exact Solutions of Two Point Time Correlations on Point Reactor Kinetics with Gaussian Reactivity Fluctuations

Station blackout is reported to be a sequence that would likely be a significant contributor to the accident risk at a boiling water reactor (BWR). The occurrence frequency of station blackout is evaluated in probabilistic safety assessment (PSA) to be 6×10^?6 per reactor year at Limerick and less than 10^?7 per reactor year at BWR in Japan.

This report describes an analytical study of thermal-hydraulic and radionuclide behavior during a postulated severe accident of station blackout at a reference BWR plant. The analytical approach was shown in both of hand calculation and the THALES/ART code calculation to better understand wide physical and chemical phenomena in the processes of severe accidents.

We evaluated timing of key events, core cooling and core temperature, reactor vessel failure, debris temperature, containment pressure, and release and deposition of radionuclide in the containment. The THALES and CORCON models on the chemical reactions in the core-concrete interaction lead to great differences in the increasing rate of containment pressure and the release rate of fission products from the core debris.     

Evaluation of Containment Failure Probability by Ex-Vessel Steam Explosion in Japanese LWR Plants

The containment failure probability due to ex-vessel steam explosions was evaluated for Japanese BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with a Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. We evaluated the conditional containment failure probability (CCFP) based on the preconditions of failure of molten core retention within the reactor vessel, relocation of the core melt into the water pool without significant interference, and a strong triggering at the time of maximum premixed mass. The obtained mean and median values of the CCPF were 6.4x 10^?2 (mean) and 3.9x 10^?2 (median) for the BWR suppression pool case, 2.2x10^?3 (mean) and 2.8x10^?10 (median) for the BWR pedestal case, and 6.8X10^?2 (mean) and 1.4x10^?2 (median) for the PWR cavity case. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for the above-mentioned preconditions. Thus, the CCFPs per core damage should be lower than the values given above. The specific values of the probability were most dependent on the assumed range of melt flow rate and fragility curve that involved conservatism and uncertainty due to simplified scenarios and limited information.     

Simplified Method for Calculation of Neutron Capture Transformation Effects of Fission Products on Decay Power

A simplified method is proposed for the calculation of the effects of neutron capture transformations of fission products (FPs) on the decay power of FPs. The decay power of FPs after shutdown changes by the neutron capture transformations of FP nuclides during reactor operation. It is proposed to calculate the neutron capture transformation effects considering the production of the following 7 nuclides ¹⁰³Ru, ¹³⁴Cs, ¹³⁶Cs, ^148mPm, ¹⁴⁸Pm, ¹⁵⁴Eu and ¹⁵⁶Eu by the neutron capture reaction of the direct mother nuclide alone giving a cumulative fission yield for the mother nuclide. The present method was assessed by com-paring the calculation results with the rigorous calculation results for the thermal-neutron fission of ²³⁵U irradiated between 1 and 5 yr in a light water reactor with thermal-nentron flux between 3 x 10¹³ and 6 x 10¹³ n/cm²·s and for the fast-neutron fission of ²³⁹Pu irradiated between 1 and 5 yr in a fast breeder reactor with total neutron flux between 3 x 10¹⁵ and 6 x 10¹⁵ n/cm²·s. It has been clarified that the present method can calculate the neutron capture transformation effects within the accuracy of ±1% of the decay power for the irradiation of 1yr and cooling time less than 10⁹s irrespective of fission type and neutron flux. The accuracy varies little with neutron flux but considerably with irradiation time. For a irradiation of 5 yr the present method can calculate the capture effect within the accuracy of +1% and -5% of the decay power. The accuracy can be improved to ±1% of the decay power with the simple correction factors.     

Status of the compact 1 MV AMS facility at the Centro Nacional de Aceleradores (Spain)

Since February 2006, the new 1 MV multielement compact AMS facility SARA (Spanish Accelerator for Radionuclides Analyses) at the Centro Nacional de Aceleradores (CNA) in Sevilla (Spain) is fully operative. During the first one and a half year of operation, the viability of the system for the measurement of ¹⁰Be, ¹⁴C, ¹²⁹I and plutonium isotopes, ²³⁹Pu and ²⁴⁰Pu, has been evaluated. First results have demonstrated that, in terms of precision and detection limits, the performance of the device compares to other compact AMS facilities, although some progress can still be done in order to optimize its capacities. At this moment, background levels are in the order of 10⁻¹⁴ for ¹⁰Be/⁹Be, 10⁻¹³ for ¹²⁹I/¹²⁷I, 10⁻¹⁵ for ¹⁴C/¹²C (processed and unprocessed blank) and about 10⁶ atoms for plutonium isotopes: ²³⁹Pu, ²⁴⁰Pu and ²⁴²Pu. In this work, the current status of the AMS measurements at CNA for the above mentioned radionuclides is described.     

Fusion-fission hybrid design with analysis of direct enrichment and non-proliferation features (the SOLASE-H study)

The role of a fusion-fission hybrid in the context of a nuclear economy with and without reprocessing is examined. An inertial confinement fusion driver is assumed and a consistent set of reactor parameters are developed. The form of the driver is not critical, however, to the general concepts. The use of the hybrid as a fuel factory within a secured fuel production and reprocessing center is considered. Either the hybrid or a low power fission reactor can be used to mildly irradiate fuel prior to shipment to offsite reactors thereby rendering the fuel resistant to diversion. A simplified economic analysis indicates a hybrid providing fuel to 10 fission reactors of equal thermal power is insensitive to the recirculating power fraction provided reprocessing is permitted. If reprocessing is not allowed, the hybrid can be used to directly enrich light water reactor fuel bundles fabricated initially from fertile fuel (either ThO₂ or ²³⁸UO₂). A detailed neutronic analysis indicates such direct enrichment is feasible but the support ratio for ²³³U or ²³⁹Pu production is only 2, making such an approach highly sensitive to the hybrid cost. The hybrid would have to produce considerable net power for economic feasibility in this case. Inertial confinement fusion performance requirements for hybrid application are also examined and an integrated design, SOLASE-H, is described based upon the direct enrichment concept.     

Sorption Behavior of Selenium(—II) on Rocks under Reducing Conditions

The sorption of radionuclides on rocks is an important factor for performance assessment of geologic disposal of radioactive wastes. Batch sorption experiments under reducing conditions were carried out to investigate the sorption behavior of selenium onto granodiorite, sandy mudstone, tuffaceous sandstone, and their major constituent minerals and accessory minerals. Rock cores and groundwater employed in the experiments were sampled with special care to avoid exposing the cores and groundwater to air to minimize their oxidation. Selenium was spiked as HSe^? and Se₄ ^2- in the experimental solutions, and reducing conditions were maintained throughout the sorption periods. Distribution coefficients, Kd (m³ kg^?1), were obtained in the ranges of 2.2 × 10^?4 to 4.0 × 10^?3 m³ kg^?1 for granodiorite, 3.3 × 10^?2 to 5.6 × 10^?2 m³ kg^?1 for sandy mudstone, and 2.9 × 10^?2 to 8.2 × 10^?2 m³ kg^?1 for tuffaceous sandstone at pHs 8.5 to 11.5. The dominant sorbent minerals for selenium over a neutral to alkaline pH range were determined to be biotite for the granodiorite samples and pyrite for the sandy mudstone and tuffaceous sandstone samples.