Reprocessing Plant Research Articles

Development of a portableα-particle counter for public acceptance activities

In order to better understand nuclear energy, the radiation counter for public acceptance (PA) activities are now widely used. The radiation counter which detects γ-rays and β-rays, respectively, has been developed and supplied commercially, but there is no counter for PA use which detects α-rays. Plutonium has been becoming the center of public interest in regard to transportation by the shipAkatsuki-maru, Rokkasho Reprocessing Plant, and the prototype FBR (fast breeder reactor) “Monju.” People in general are very concerned about plutonium, which emits α-rays, because of its toxicity and possible use in nuclear weapons. In this connection, we have decided to develop a portable α-particle counter for PA activities which detects α-rays easily. We use Si surface barrier type semiconductor as a detector, and the total system is compact, light in weight, easy to handle, and inexpensive. Four kinds of samples are prepared; soil gathered from a hot spring field, sinter (siliceous material deposited by springs), uranium ore, and monazite. People in general can easily understand the nature of α-rays which exits around us.

The Thorp Project - An Overview

BNFL is an international company offering a nuclear fuel service. BNFL owns and operates facilities for the storage and reprocessing of irradiated fuel, and treatment of wastes arising, at the Sellafield site in West Cumbria. In 1974, BNFL announced its intention to undertake the Company's largest ever project, the provision of a new, integrated, reprocessing facility known as THORP (Thermal Oxide Reprocessing Plant). The purpose of THORP is to recover uranium and plutonium from spent oxide fuel that has been irradiated in nuclear reactors used for the generation of electricity. The plant has been designed to high standards to avoid jeopardising the safety of any person on or off site as a result of its operation. This paper provides an overview of the project outlining some of the major aspects, encompassing the history of the project, environmental impact, safeguards/accountancy, commercial information, the use of the products in mixed oxide fuel and the development of the THORP workforce. It concludes that the large investment made in plant, equipment and people, will ensure that the radiological impact of THORP's operations on the environment is insignificant, and that as the radioactive commissioning of THORP is proceeding successfully, that there is increasing confidence within, and external to, BNFL that THORP will be a commercial and environmental success for Britain.

The Application of Quality Management Systems for the Handling, Loading and Transport of Spent Nuclear Fuel from German Reactors to European Reprocessing Plants

AbstractThe tasks carried out by NTL-Hanau, which has been working in the field of transport of radioactive material for more than 20 years, are described. The NTL Group procedures, which were drafted nearly 10 years ago, have been developed into a comprehensive Quality Assurance system. This paper describes the QA system following ISO 9001 and IAEA Safety Series No 37 but which, for German QA requirements, has also to conform to German KTA Standard No 1401. The preparation of shipments includes, inter alia, the planning of the transport, flask utilisation, the provision of transport vehicles and the assignment of personnel. The flask handling procedures are also described, including the essential QA step of identifying fuel elements before loading. NTL-Hanau generally acts as the consignor of each shipment and this entails the provision of services such as the performance of final flask checks, completion of transport documentation and proper labelling of the package. In addition NTL-Hanau is continuousl...

Burnup Credit in Spent Fuel Transport to Cogema LA Hague Reprocessing Plant

Among the spent fuel elements received yearly at the COGEMA La Hague reprocessing plant, an ever growing number benefit from burnup credit for transport or storage purposes. The burnup credit level is determined by the package or storage rack criticality assessment and approved by the French safety authorities. Depending on this level, the fuel irradiation is controlled either by qualitative or quantitative measuring techniques. A description of the documentary and operational system developed by COGEMA and NTL for fuel identification and burnup control is presented.

Effects of Alpha Decay on the Properties of Actual Nuclear Waste Glass

AbstractAn accelerated experiment using the actinide doping technique was performed to investigate the effects of alpha decay on the properties of actual nuclear waste glass at high radiation doses. A fully radioactive borosilicate waste glass, containing the actual high-level radioactive liquid waste generated from the Tokai Reprocessing Plant of PNC, was prepared by JAERI, and a powder mixture of the ground fully radioactive glass and 244CmO2 was melted at 1200°C for 2 hrs. The radioactivity concentration of 244Cm was 1.0 } 1010 Bq/g-glass at the date of preparation. The homogeneity of curium-doped glass samples was confirmed by the density measurement, heat load measurement and alpha autoradiography. The properties of the irradiated samples were investigated by the mass spectrometer for helium determination, the optical microscope, the electron probe microanalyzer, the densitometer, the Soxhlet and MCC-1 leach testing apparatus. By measuring the amount of helium released from the curium-doped glass samples, more than 99% of helium remained in the matrix at room temperature. The density of the sample slightly decreased with the increase of cumulative alpha decays and the decrease of 0.77% was observed at a dose of 1.55 } 1018 alpha decays/g, corresponding to an equivalent age of 150000 yrs. Optical and scanning electron micrographs showed that no cracks were observed on all samples having up to a dose of 1.5 } 1018 alpha decays/g. The leach rates, based on weight loss, in both the Soxhlet (100°C, 7 days) and MCC-1 (90°C, 28 days) tests did not significantly change with alpha decay dose.

An application of automated operating procedure synthesis in the nuclear industry

The synthesis of operating procedures for the transient operation of chemical processes is a formidable and expensive task during the design of a plant. It is time-consuming and error-prone both for the process engineering department to specify the procedures and for the automation department to generate sequence control code from the specifications. A further source of error arises at the interface between the two departments since the boundary is typically ill-defined and engineers of different backgrounds often make assumptions without adequate understanding of the other discipline. In this paper, we describe the application of the CAPS (computer-aided procedure synthesis) system (Crooks et al., 1992a,b,c) to a case study in the nuclear industry — British Nuclear Fuels' (BNFL) Thermal Oxide Reprocessing Plant (THORP), which will be the largest facility in the world for reprocessing spent reactor fuel pins. The case study considers the “head end” plant of THORP which involves reagent preparation and fuel pin dissolution and is a single product batch and semi-continuous process. The CAPS system successfully generated realistic supervisory procedures and control sequence specifications which were in close agreement with the actual control sequences developed by BNFL engineers. Furthermore, the modelling effort required as input to CAPS was reasonable and all of the information was readily available early in the design life cycle.

六ケ所再処理施設の安全設計と安全評価

六ケ所再処理施設の安全設計と安全評価

Environmental Dose Assessment For Low-level Radioactive Effluents Discharged From Tokai Reprocessing Plant

In the normal operation of the Tokai reprocessing plant, low-level radioactive effluents are discharged to the atmosphere and the ocean under rigid control. Radiation exposures to the population around the plant have been estimated for potential pathways with site-specific parameters such as food consumption, concentration factors of marine organisms, and meteorological conditions. External exposures to a radioactive cloud and internal exposures from inhalation and oral intake of radionuclides are evaluated for airborne effluents. External exposures to a contaminated fishing net and fishing boat are considered pathways for fishermen. External exposure to a contaminated beach and internal exposure via oral intake of marine products are evaluated for liquid effluents. Since the plant began operation in 1977, estimated effective dose equivalents for the public have been less than the annual effective dose equivalent limit of 0.1% recommended by the International Commission on Radiological Protection (1977).

The importance of soil adhered to vegetation as a source of radionuclides ingested by grazing animals

Soil ingestion has been identified as a potentially important source of radionuclides to grazing animals. Seasonal patterns of soil adherence to vegetation and its implications for the radionuclide intake of grazing animals were measured at two west Cumbrian sites. Soil adhesion to vegetation was highly seasonal, being highest in autumn and winter. At Site 1 (a lowland pasture close to the British Nuclear Fuels plc Sellafield Reprocessing Plant, Cumbria, UK), vegetation samples were found to consist of up to 46% soil (by dry weight). Therefore, the importance of soil as a potential source of radionuclides to grazing animals was also seasonal; soil comprised up to 92% of the 137Cs and potentially all of the 239 240 Pu of vegetation samples at Site 1 and up to 62% of the 137Cs at Site 2 (an upland farm contaminated following the Chernobyl accident). Analyses of sheep faecal samples confirmed the seasonal importance of soil as a potential source of radionuclides. The importance of soil adhering to vegetation, as a source of contaminant 137Cs, increased with time after the Chernobly accident. Calculations of soil ingestion must be made when the intake of radionuclides by grazing animals is being studied. In some circumstances, present radioecological models may considerably underestimate the importance of soil ingestion.

Determination of Technetium in Spent Fuel Reprocessing Process Streams by Glove-Box-Enclosed Inductively Coupled Plasma Emission Spectrometer

This report describes the determination of technetium in solutions from spent fuel reprocessing processes. Inductively coupled plasma emission spectrometry (ICPES) was employed after alkaline precipitation and/or anion exchange. Technetium in the form of pertechnetate is roughly separated from other elements by merely changing the acidity of the sample solution to make it alkaline. This works because most of the fission products (FPs) are precipitated alongwith the heavy metals in an alkaline medium, whereas technetium is not. Technetium in some solutions from highly active processes is further purified with the aid of basic or weakly acid anion exchange after the alkaline precipitation mentioned above. The decontamination factors were found to be more than 104 and 105 for FPs and for the heavy metals, respectively. The amount of technetium separated is determined by ICPES, which measures the Tc II 261.00nm line without any significant interference. The relative standard deviations are 2.2% at 9μg ml-1 and 4.9% at 0.9μg ml-1 of technetium. The proposed method has been applied to samples from various streams of the Tokai Reprocessing Plant; the highly active liquid waste, the spent fuel dissolver solution, the aqueous phase of solvent extraction and the plutonium product solution.

An Inference Model for Predicting a Pinching Effect in the Co-Decontamination Extraction Process in a Purex Fuel Reprocessing Plant

An Inference Model for Predicting a Pinching Effect in the Co-Decontamination Extraction Process in a Purex Fuel Reprocessing Plant

Iodine-129 in the environment of a nuclear fuel reprocessing plant: VII. Concentrations and chemical forms of 129I and 127I in the atmosphere

Gaseous and particulate iodine samples have been collected simultaneously from the atmosphere around the Karlsruhe Reprocessing Plant (WAK) using a three-component sampler which separates the iodine into three fractions, namely: (a) associated with aerosol particles, (b) gaseous inorganic and (c) gaseous organic. These field measurements were conducted at five sampling sites located between 0·7 and 23·0 km from the WAK exhaust air stack in the dominant downwind direction. In 1986 the air samples were collected between October and December and in 1987 between June and September. The aerosol fraction of 129I in 1986 ranged between 5·2 and 30·8% and in 1987 between 1·7 and 9·7%. The corresponding values for 127I were 12·9 and 28·0% (1986) and 16·4 and 9·4% (1987). The gaseous inorganic 129I fraction varied between 26·8 and 35·2% (1986) and between 17·2 and 27·3% (1987). Corresponding 127I data were: 1·6 and 26·2% (1986) and 18·7 and 26·6% (1987). The gaseous organic fraction of 129I accounted for 34·0–65·9% (1986) and 62·9–88·8% (1987). Corresponding 127I values were: 46·1–73·8% (1986) and 59·3–71·9% (1987). A corresponding data set from a ‘control’ site in Bonn gave the following results: aerosol fraction: 129I: (6·0 ± 1·0) × 10 −16 g/m 3; 127I: (2·0 ± 1·9) × 10 −9 g/m 3; gaseous inorganic iodine: 129I: (1·1 ± 4·2) × 10 −15 g/m 3; 127I: (1·5 ± 0·9) × 10 −9 g/m 3; gaseous organic iodine: 129I: not detectable; 127I: (8·7 ± 3·5) × 10 −10 g/m 3.

Long-Term Monitoring of Workers at the Cogema Fuel Reprocessing Plants

Long-Term Monitoring of Workers at the Cogema Fuel Reprocessing Plants

Development of Microwave Heating Method for Co-Conversion of Plutonium-Uranium Nitrate to MOX Powder

The development project of a conversion process using a microwave heating method (MH method) started in 1977. After basic experimental works, a co-conversion test unit was installed in the Plutonium Fuel Fabrication Facility (PFFF), PNC in 1979. The test unit had been operated to get engineering data of the MH method until 1984. During the operation of the test unit, the evaluation of the process and the characterization of the products powder have been performed. The data shows that MH process has many excellent advantages, such as good characteristics of the MOX product powder, good homogeneity of Pu-U in pellets, simplicity of the process, and a little liquid waste etc. Based on these data, the Plutonium Conversion Development Facility (PCDF) with the capacity of 10 kg Pu+U/d was designed and constructed beside the Tokai Reprocessing Plant. After a test operation in 1983, the PCDF has been operated quite smoothly without trouble. The products MOX powder has been supplied as a raw materials of fuels for the ATR “FUGEN” and the FBR “JOYO”. Total quantity of the converted MOX powder was attained to about 4 t and over 50 t of MOX fuels have been fabricated and shipped to the reactor sites until July of 1988. Through the operations of the test unit in the PFFF and the PCDF, so many studies on MH method have been also performed and continuous denitration units are under development for the future technology.

Long-Term Monitoring of Workers at the Cogema Fuel Reprocessing Plants

A brief history is given of occupational medicine in France and of the gradual organisation of specific monitoring systems for workers in the nuclear industry. Two factors have proved essential for effective monitoring: the use of a record file of individual work-stations and risk factors, and the conclusions drawn by a panel of experts at regular meetings held to review the results of monitoring. Principles governing routine monitoring and post-incident monitoring have been established on the basis of urine and faeces tests. The laboratory techniques used for these tests are described, as well as the principles of their interpretation, stated in two key reports by the Commissariat à l'Energie Atomique. On the basis of the data acquired from these tests, it is concluded that calculations of committed doses were only required in a very few cases relative to the number of workers monitored. Concomitantly with the biological tests, atmospheric sampling is being instituted in workshops.

A Comparison of Results of Methods of Individual Monitoring at a Fuel Reprocessing Plant

A Comparison of Results of Methods of Individual Monitoring at a Fuel Reprocessing Plant

A Comparison of Results of Methods of Individual Monitoring at a Fuel Reprocessing Plant

Abstract Until recently the method of obtaining individual assessment of fast neutron dose equivalent at BNFL Sellafield was based on occupancy studies and the use of neutron to gamma dose rate ratios. Essentially, the method entailed a measurement of neutron to gamma dose equivalent ratios at sites occupied by groups of workers, this ratio being then applied to the gamma dose equivalent recorded by personal film badge dosemeters to obtain an estimate of the neutron personal dose equivalent. Following extensive trials, both on phantoms and personnel, the NRPB PADC (CR-39) personal neutron dosemeter is now used to assess the personal fast neutron dose equivalent. A comparison has been made of the assessments made using the two methods. This comparison shows that there is agreement for average dose equivalents for groups of workers. However, there are differences for individual workers. After consideration of the variation in neutron to gamma ratios between working areas and the incompleteness of the occupancy data, the conclusion reached is that the more accurate assessment is obtained by the use of personal dosemeters.

Static Leaching of Actinides and Fission Products from Fully Radioactive Waste Glass of Hllw Generated in Tokai Reprocessing Plant

ABSTRACTFully radioactive waste glass of HLLW generated in Tokai Reprocessing Plant was produced by using a small Joule-heated ceramic melter at the Chemical Processing Facility (CPF), Tokai Works, PNC, Japan. Static leach tests using the glass were carried out in doubly distilled water for a period of up to three months at 30°C and 90 °C. Gamma spectrometry was performed to measure 134Cs, 137Cs and 60Co in the leachates. After the chemical separation of Cs isotopes in the leachates, it was newly possible to measure 125Sb, 144Ce, 154Eu and 155Eu. Alpha spectrometry was performed to measure 239Pu+240Pu, 241Am and 244Cm in the leachates. Adsorption of the actinides on the wall of the leach vessel was recognized in all cases. Leaching behavior of the measured radionuclides were evaluated by considering mass balance of the radionuclides which were leached.

The Dounreay inquiry: Public participation in practice

Abstract The Public Inquiry into the application for planning permission for the construction of the European Demonstration Fast Reactor Fuel Reprocessing Plant (EDRP) at Dounreay, Caithness, is reviewed. The roles of the different participants in the inquiry are examined, and some conclusions are drawn on both specific and general issues relating to the Inquiry.

Japan to Build French Reprocessing Plant

Japan to Build French Reprocessing Plant