Fuel Fabrication Plant Research Articles

Development of high-efficiency passive counters (HEPC) for the verification of large LEU samples

A paper describing the conceptual idea of using passive neutron assay for the verification of large size uranium samples in fuel fabrication plants was first presented at the 2001 ESARDA conference. The advantages of this technique, as a replacement of active interrogation using the PHOto-Neutron Interrogation Device (PHONID) device, were evident provided that a suitable detector with higher efficiency than those commercially available would be realised. The previous paper also included a feasibility study based on the experimental data. To implement this technique, a high-efficiency passive counter (HEPC) has been designed by the JRC, Ispra. JRC has also built a first smaller-scale prototype. This paper will describe the tests made in the PERLA laboratory and report the performance of the prototype. In parallel, the design of the large HEPC has been finalised for Euratom safeguards. Two units for the fuel fabrication plants in Dessel (B) and Juzbado (E) have been produced by a commercial manufacturer under JRC specifications. The two detectors have been installed in the two sites in summer 2004 after an extensive test campaign in PERLA. Since then they are in use and some feedback on the experience gained is reported at the end of this paper.

Application of neutron well coincidence counting for plutonium determination in mixed oxide fuel fabrication plant

Mixed oxide (MOX) fuel is an alternative to conventional enriched uranium oxide fuel in thermal reactors. Indian interest in plutonium recycle in thermal reactors is primarily due to the need to develop alternative indigenous fuel for two boiling water reactors (BWR) at Tarapur, which are designed to use imported light enriched uranium fuel. A few MOX assemblies have been fabricated and loaded into the reactors. Neutron well coincidence counting (NWCC) system has been successfully employed to check the enrichments of PuO2 in MOX blends. NWCC has also been successfully applied in developing dry recycling process of clean rejected oxide (CRO) and dirty rejected oxide (DRO).

Design and performances of the Scrap Neutron Multiplicity Counter

JRC has developed the Scrap Neutron Multiplicity Counter (SNMC): an advanced neutron multiplicity counter for the verification of inhomogeneous Pu samples, such as scrap material in MOX fuel fabrication plants. The innovative features of this counter with respect to existing ones rely on two aspects: (i) an optimised design based on Monte Carlo calculations in order to select the most appropriate materials, geometry and detector disposition for maximum efficiency and (ii) novel electronics based on digital signal processing (DSP) reducing the system dead time. The paper recalls the design process, the electronics, the construction and assembly of the counter. Then the results of the first experimental tests will be reported. We will show the characterization of the main physical parameters of the counter, the calibration and the verification of a wide variety of plutonium bearing samples available in the PERLA laboratory at JRC Ispra. This will include pure homogeneous samples (Pu dioxide powders, metal Pu, MOX powders and pellets) and some tests on heterogeneous samples representative of scrap material.

Proliferation Resistance Assessment Methodology for Nuclear Fuel Cycles

A methodology, based on the multiattribute utility analysis, for the assessment of diverse fuel cycles for proliferation resistance was developed. This methodology is intended to allow for the assessment of the effectiveness of safeguards implementation at facilities within a large-scale fuel cycle and allow for the ability to choose technologies based in part on their effectiveness to deter the proliferation of nuclear materials. Fuel cycle facilities under consideration include nuclear reactors, reprocessing facilities, fuel storage facilities, enrichment plants, fuel fabrication plants, uranium conversion plants, and uranium mining and milling operations. The method uses a series of attributes (for example, Department of Energy attractiveness level, weight fraction of even Pu isotopes, measurement uncertainty, etc.) to determine a proliferation resistance measure for each step in a process flow sheet. Each of the attributes has a weighting that determines its importance in the overall assessment. Each attribute also has an associated utility function derived from both expert knowledge and physical characteristics that relates changes in the value of the attribute to its overall effect on the proliferation resistance measure. A method for aggregating proliferation resistance values for each process in a flow sheet into an overall nuclear security measure for the complete cycle was also developed. This method is focused on preventing host nation diversion; however, a similar technique could be used to analyze the risk due to theft by an insider or outsider. This methodology has been applied successfully for example fuel cycles to demonstrate its viability as an assessment methodology and its capability in discriminating diverse fuel cycle options.

Energy and Angular Responses of the Criticality Accident Detector using a Plastic Scintillator

The Japan Atomic Energy Agency (JAEA), Nuclear Fuel Cycle Engineering Laboratories, operates a spent fuel reprocessing plant and MOX (Plutonium-Uranium Mixed Oxide) fuel fabrication plants. Criticality accident detectors have been installed in these facilities. The detector, the Toshiba RD120, is composed of a plastic scintillator coupled to a photomultiplier tube, and an operational amplifier. The alarm triggering point is set to 1.0–3.6 mGy.h−1 in photon dose rate to detect the minimum accident of concern. However, a plastic scintillator is principally sensitive not only to primary photons but also to neutrons by secondary photons and heavy charged particles produced in the detector itself. The authors calculated energy and angular responses of the RD120 criticality accident detector to photons and neutrons using Monte Carlo computer codes. The response to primary photons was evaluated with the MCNP-4B and EGS4 calculations, and photon and X-ray irradiation experiments. The response to neutrons that produce secondary photons and heavy charged particles from neutron interactions was computed using the MCNP-4B and SCINFUL, respectively. As a result, reliable response functions were obtained. These results will be a great help in reassessing the coverage area and in determining the appropriate triggering dose rate level in criticality accidents.

Micro- and meso-scale variability in sediment activity concentrations: The implications for dose assessment

Radionuclides in the environment require regular monitoring to establish activity concentrations and make reliable dose estimates. This study examines the variability in selected radionuclide activity concentrations at a variety of spatial scales and assesses the resulting errors in dose estimates. The Ribble Estuary, UK provides an ideal location to conduct non-destructive tests (β -dose rate measurements) due to the discharge of 234 Th and Pa from the BNFL fuel fabrication plant at Springfields (Lancashire, UK). This estuary also receives Sellafield derived radionuclides after transport through the Irish Sea. Results indicate that significant errors in external beta and gamma dose assessments may result from the use of single sediment activity concentration or dose values to represent an area of contaminated sediment. For example, in an area only 1m × 1m square, a field coefficient of variation for beta dose may be as high as 25%, even when there are no visible textural or compositional variations to account for this, and this increased to 83% at the 100×100m scale. Erroneous conclusions on the nature of change in time-series measurements can be made if local field variability has not been assessed and allowed for when planning the monitoring programme.

Decommissioning and dismantling of the Siemens fuel cycle facilities

Abstract Four Siemens-owned German fuel cycle facilities are being decommissioned at the Hanau and Karlstein sites. Decommissioning of one of these facilities has already been completed. The facilities comprise three fuel fabrication plants and a facility for post-irradiation investigations of fuel assemblies. In each case the actual decommissioning phase was preceded by clean-out. Extensive preparatory investigations and planning, including licensing procedures, had been necessary. In some cases new measuring techniques and systems had to be developed and waste processing systems erected. In addition to providing a general survey of the different conditions and procedures involved, the MOX facility is used in this paper as an example for presenting these activities in greater detail. This includes the concept of immediate conditioning of radwaste for maintenance-free long-term interim storage (storage for several decades until a repository is available) in such a way as to also meet the requirements for shipment and final disposal without any need for further actions. A number of lessons have been learned and recommendations derived from the many years of decommissioning work at the four facilities.

Experience in Japan in Adopting IAEA ST-1, 1996 Edition

Since Japan does not have rich energy resources, it is essential for Japan to ensure energy security in order to sustain economic growth and social activities. To secure a stable energy supply, the establishment of nuclear fuel recycling has a vital role in the use and development of nuclear power. Safe and smooth transport between the various nuclear fuel recycling facilities, enrichment plants, re-conversion plants, fuel fabrication plants, reprocessing plants, etc. is therefore essential and of increasing importance in the nuclear fuel cycle field. In the Japanese regulatory system, radioactive material is classified into radioactive isotopes and nuclear fuel material (which includes nuclear raw material). These two categories are regulated by different regulations. Radioactive material is widely utilised to improve the quality of plants, to preserve food and to make medical examinations etc. Nuclear fuel material is mainly utilised as fuel for nuclear power plant and research reactors.

Comparison of neutron dose quantities and instrument and dosemeter readings at representative locations in an MOX fuel fabrication plant

The relationships between operational and protection quantities, and values of personal dosemeter and instrument readings have been determined for a recently designed MOX fuel fabrication plant. The relationships between the quantities, and the readings of personal dosemeters are sensitive to both the energy and direction distribution of neutron fluence. The energy distributions were calculated using the Monte Carlo code MCBEND. The direction distribution was addressed by calculating independently, spectral components for which the direction distribution could be reasonably assumed. At representative locations, and for assumed worker orientations, the radiation field is analysed as having, in general, three components—a direct, unidirectional component from the nearest identified discrete source, which is considered incident A-P, several unidirectional components from other such sources which are treated as a rotational component and a scattered isotropic component. The calculated spectra were folded with conversion coefficients for personal dose equivalent, H p(10) slab (A-P, ROT and ISO), effective dose, E, (A-P, ROT and ISO), ambient dose equivalent, H *(10), personal dosemeter (AP, ROT and ISO) and survey instrument response characteristics.

Airborne uranium contamination — as revealed through elemental and isotopic analysis of tree bark

A new strategy for characterisation of airborne uranium contamination based on ICP mass spectrometric analysis of tree bark is described. The uranium content of tree barks (50 samples) obtained from diverse locations (remote, rural, industrial) varied over almost four orders of magnitude (0.001–8.3 μg/g U) with maximum concentrations recorded in the vicinity of a nuclear fuel fabrication plant (0.70–8.3 μg/g U). Elevated concentrations were also observed near a coal-fired power station (0.25–0.38 μg/g U). Isotopic analysis revealed significant deviation from the natural uranium isotope ratio (235U/238U, 0.00725) at four nuclear installations (235U/238U, 0.0055–0.0097). These findings indicate that tree bark serves as an effective biomonitor for uranium and, with isotopic analysis, discrimination between nuclear and non-nuclear emissions is realised.

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors—I: DUPIC Fuel Fabrication Cost

A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (~49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC.

Characterisation of airborne uranium and thorium contamination in Northern England through measurement of U, Th and 235U/238U in tree bark

Samples of tree bark were collected from four locations in Northern England (a typical rural site, a coal-fired power station, a uranium (isotopic) enrichment plant and a nuclear fuel fabrication facility), to assess the nature and extent of airborne uranium and thorium contamination. The U and Th concentrations of bark were determined by inductively coupled plasma mass spectrometry after conventional nebulisation of bark digests, whilst measurement of 235U/238U isotopic ratio utilised high efficiency nebulisation. Uranium concentrations varied between and within the sites (range, 0.01-12 micrograms g-1), with maximum values recorded within 1 km of the nuclear fuel fabrication plant (Springfields). In comparison, the concentration of Th in bark was low (mean, 0.018 microgram g-1) at all sites with the exception of the area affected by coal combustion (0.2-0.8 microgram g-1). The U/Th ratio varied from 0.5 to 3900 compared with the average crustal ratio of 0.3. Low values (< 2) were recorded at the 'coal' and 'rural' sites whilst Capenhurst and Springfields showed high values indicating the relative magnitude of uranium elevation. Significant enrichment of the natural 235U/238U ratio (0.00725) was observed near the nuclear installations, in particular, the enrichment plant (Capenhurst).

Determination of neutron spectra in a MOX plant for the qualification of the BD-PND bubble detector.

As a result of the introduction of the ICRP 60 recommendations and the increasing contribution of the neutron dose to the total dose of the personnel at the Belgonucleaire Mox fuel fabrication plant, the BD-PND bubble detector manufactured by Bubble Technology industries was introduced as a new, reliable personal neutron dosimeter. In the framework of the evaluation program of the bubble detector, measurements and calculations of the neutron spectra in the installations of the fuel fabrication plant were performed. The measurements were carried out with a ROSPEC neutron spectrometer, and the calculations were performed by means of the Monte Carlo code MCNP 4A. Comparison between measurements and calculations revealed good agreement. On the basis of the obtained neutron spectra, a correction factor was determined to take into account the new ICRP 60 recommendations and the difference between the calibration spectrum of the bubble detectors and the observed neutron spectra at the plant. This correction factor was applied to the calibration factor provided by Bubble Technology Industries.

Fiploc analysis of an accident with uranium-aerosol in a German fuel fabrication plant

Fiploc analysis of an accident with uranium-aerosol in a German fuel fabrication plant

Reactors-related utilization options of weapons plutonium being released from the defence program

The release of considerable amounts of fissile materials has once again drawn the attention of politicians, scientists and the public to the nuclear problem, as has often been the case in the past when nuclear arms were being developed and tested. An essentially new factor which makes the current status of relations different from the former ‘cold war’ years is the spirit of openness and collaboration between states in response to this problem. It is known that Russia and the United States have differing viewpoints concerning the most dangerous material—plutonium, though recently there has been a perceptible convergence of their respective positions. Russia from the very beginning has planned the separation of civil plutonium from the spent nuclear fuel and utilization of the former in nuclear reactors for electricity production. In this case the handling of surplus weapons plutonium can be considered as an additional episode in the nuclear power development strategy. In the United States, on the contrary, following the rejection of the policy of separating and recycling civil plutonium, the surplus weapons plutonium handling program is being considered, as a matter of fact, separately from the national nuclear power program. That is why burial in geological formations has been considered for a long time as the only principal method for disposal of surplus weapons plutonium. Nevertheless, recently, to a considerable degree thanks to the position of Russian scientists, the United States has had to recognize the expedience of utilization of weapons plutonium as a mixed oxide (MOX) fuel for light water reactors, in addition to the burial option. Presently in Russia various ways of surplus weapons plutonium utilization in fast and in thermal reactors are being considered. From the practical point of view, the most important question for each of the options is the amount of weapons plutonium utilized and the respective costs ratio. Preliminary analysis has shown that the extent of costs for weapons plutonium utilization for any of the options depends upon employment of new or partially constructed installations. In cases where only existing or to be upgraded installations are used, the costs look real enough. Employment of new or partially constructed installations will increase the total cost on principle to a new level. This factor is especially important for reactor variants of plutonium management, which include, as a minimum, two installations—the MOX fuel fabrication plant and the nuclear reactor itself.

ＴＲＵ廃棄物処分場への人間侵入の観点から見た地下構造物深度分布調査

Depth distributions of some kinds of underground structure in Japan have been investigated to get an information about suitable depth of underground repository for TRU waste that is arising from reprocessing and MOX fuel fabrication plants. The underground structures investigated in this work were foundation pile of multistoried building, that of elevated expressway, that of JR shinkansen railway, tunnel of subway and wells. The major depth distribution of the underground structures except for the wells was in range from 30 to 50m, and their maximum depth was less than 100m. On the other hand, the 99% of wells was less than 300m in depth. Maximum depth of the other underground structures has been also investigated for a survey of the utilization of underground by artificial structures in Japan.

Assay of uranium in U-bearing waste produced at natural uranium metal fuel fabrication plants by gamma-ray spectrometry

A passive gamma measurement technique is investigated for the assay of U-bearing wastes generated at the Natural Uranium Metal Fuel Fabrication Plants. A 3 in. × 3 in. NaI(Tl) detector was used in conjunction with a multichannel analyzer (MCA). The observed count rate of the 1 MeV γ-ray from the 238U in the sample was corrected for sample self-absorption and for absorption in the walls of the sample container. These correction factors were determined using one reference standard and from a knowledge of the sample weight, composition and the geometry of the sample container made of pure aluminium. The amount of 238U in the samples were measured by comparing the corrected area under the 1 MeV gamma ray peak of the known reference standard with the corresponding corrected peak areas of the samples to be measured. To compare the nondestructive assay (NDA) results with another independent method, chemical analysis of all the U-bearing waste samples was also carried out. The NDA results were found to agree within ±15% with the chemical analysis results. To make the method cost-effective, rapid and useful for plant operation on a routine basis, all the measurements were repeated using a single channel analyzer (SCA) also. The NDA results obtained with SCA agree within ±30% with the chemical analysis results.

Modelling and measurement of the dispersion of radioactive emissions from a nuclear fuel fabrication plant in the U.K.

The ground-level air concentrations of supported and total 234Th were determined over a period of 1 year at four sites in the vicinity of a nuclear fuel fabrication plant in northwest England. Both supported and unsupported 234Th were found, with mean monthly concentrations of total 234Th 13 times those observed at a control site 50 km north of the factory. Bulk deposition rates of total 234Th were found to be 2–3 times the background values. Analysis of mosses revealed a systematic decrease in supported 234Th concentrations with distance from the factory, with background values reached at a distance of about 1.5 km. Road dust samples showed a similar decrease in supported 234Th concentrations with distance, but with significant departures from this trend, possibly being evidence for the mechanical transport of contaminated dust from the factory. A Gaussian plume dispersion model was able to predict 85% of the observed air concentrations within a factor of four, despite large uncertainties in some source parameters.

Determination of Low Activities Released by a Decommissioned Fuel Fabrication Plant Through Natural Air Exchange

A method to determine the release of low activities through natural air exchange from a decommissioned fuel fabrication plant is described. The method has been applied to the buildings of the NUKEM-A plant and was important in obtaining governmental authorization for the plant decommissioning. The air exchange rate in the NUKEM-A plant was measured by using a tracer gas method. For that purpose, N2O as inert gas was injected into representative rooms, and the decrease of concentration caused by exfiltration processes was measured by an infrared gas analyzer as a function of time. The knowledge of this decay curve allows the calculation of low activities, which may be released into the environment by the natural air exchange. The activity is determined according to the German radiation protection regulation. From this, an air exchange rate of ∼25 h−1 would be equivalent to 10% of the tolerable activity emission. The measured exchange rates are less than ∼0.5 h−1. This results—at least for the meteorological conditions during the measurement period—in a significantly lower activity release than that permitted. The measuring method was successfully performed and can be recommended for similar investigations.

核燃料施設におけるハンドリング技術

核燃料施設におけるハンドリング技術