Mixed Oxide Fuel Fabrication Facility Research Articles

A Case Study for Overcoming Obstacles During Planning and Construction Phases: MOX Fuel Fabrication Facility

The decision to construct the MOX Fuel Fabrication Facility in Aiken County, South Carolina, was largely a result of the Plutonium Management and Disposition Agreement between the United States and the Russian Federation, signed in September of 2000. However, the construction of the MOX Fuel Fabrication Facility has proven to be a more complex endeavor than many of the project planners had initially anticipated. To date, the project has yet to be completed, and in February 2019, the National Nuclear Security Administration issued a contract termination notice for the project. Currently, the United States Department of Energy is looking into what it refers to as the dilute and dispose approach to dealing with the nation’s aging stockpile of weapons-grade nuclear material. The present research aims to explore solutions for overcoming the failures that occurred during the planning and construction phases of the MOX Fuel Fabrication Facility. The paper proposes that the adoption of Integrated Project Delivery (IPD) and Public-Private Partnership (PPP) might provide solutions for such cases. This study is supported by a review of relevant literature examining the effects of implementing both IPD and PPP. The results of the study represent discussions and a summary of the research contribution.

Calculated Critical and Subcritical Data for Revision of ANSI/ANS 8.12

Calculated critical and subcritical data for systems of mixed oxides of plutonium and uranium are presented for review and subsequent incorporation into a revision of ANSI/ANS 8.12. The system specifications were prepared in parallel with the ISO standard ISO 11311:2011, involving experience from international mixed oxide fuel fabrication facilities. Data are presented for two mass fractions of plutonium to uranium plus plutonium (at 12.5% and 35%), damp mixed oxide density up to a theoretical density of 11.03 g/cm3 (for the 12.5% mass fraction, ignoring the variation due to plutonium isotopic content) and up to a density of 3.5 g/cm3 (for the 35% mass fraction), three plutonium isotopic mass fractions (100% 239Pu, 95% 239Pu with 5% 240Pu, and a reactor grade composition with 20% 240Pu and the balance 239Pu, 241Pu, and 242Pu), systems at two moderation conditions (water mass fractions at 3% and at optimal conditions) and systems at two water reflector conditions (nominal 2.5 cm and full 30.0 cm). Parameters of interest presented are values of volume (L), infinite cylinder diameter (cm), infinite slab thickness (cm), mass (kg), cylinder linear density (g/cm), and slab areal density (g/cm2). Values are presented at calculated multiplication factors of 1.000, 0.980 and 0.950. Calculations were made by various subject matter experts with various computational programs and cross section libraries. Validation methods were reviewed and discussed herein though data propagated to revision of ANSI/ANS 8.12 will reflect a conservative subcritical margin selected based on expert judgment.

Plutonium management policy in the United Kingdom: The need for a dual track strategy

The United Kingdom holds the largest stockpile of separated civil plutonium in the world, projected to reach 140t, at the end of this decade, when reprocessing operations are complete. UK Government policy is that this material should be reused as MOX fuel in Light Water Reactors. This policy is re-examined in the light of recent experience of the US plutonium disposition programme, in which the MOX Fuel Fabrication Facility is now considered to be potentially unaffordable. Problematic aspects of US programme, relevant to the UK scenario, are reviewed, to understand the possible impact on UK policy. Based on the US experience and inherent uncertainty regarding the capital and operational costs of MOX fuel fabrication and plutonium immobilisation facilities, and the associated technical risks, it is concluded that the UK policy should explicitly adopt a dual track strategy to plutonium management, with commitment that: any remaining plutonium which is not converted into MOX fuel, or otherwise reused, will be immobilised and treated as waste for disposal. This will also ensure that the UK is positioned and prepared to take forward an immobilisation and disposal programme for the plutonium stockpile, should reuse as MOX fuel not prove an economic or viable option.

Neutron spectral and dose distribution studies during fast reactor fuel fabrication

In mixed oxide (MOX) fuel fabrication facilities, the fabrication of MOX fuel involves various metallurgical operations such as mixing and milling of weighed quantities of uranium and plutonium oxides, precompaction, granulation of precompacts, final compaction of granules, sintering of green pellets, followed by fuel pin fabrication. All these operations are carried out in glove boxes, which have complex geometries due to housing of various equipments. Plutonium, a source of neutrons, is handled in large quantities in various forms such as powder, granules, pellets and pellets encapsulated in pins during these operations. It is obvious that extensive knowledge on the neutron spectral distributions in workplace is required from radiation protection and shielding point of view. In this paper, a brief introduction to the source of neutrons in MOX fuel handling facilities, studies that include experimental measurements of neutron spectra of various forms of MOX, contribution of neutron fluence in various energy groups and its dose equivalent, establishment of a simulation procedure for glove boxes handling Pu using FLUKA Monte Carlo codes, comparison of simulation results with the actual experimental measurements are presented. The results indicate that bare PuO 2 and MOX sources require 3 mm of lead shield to eliminate gamma interference in MICROSPEC with N-probe neutron spectrometer. Furthermore, the studies reveal that dose equivalent contribution from MOX pellets is very significant in the energy group of 1.0-1.5 MeV unlike PuO 2 powder and fuel pins, which exhibit significant contribution in the energy groups of both 1.0-1.5 MeV and 2.0-3.0 MeV. Also, the fuel pins show high neutron fluence in the energy group of 0.0-0.01 MeV, but they do not contribute significantly to dose equivalent. A good agreement between experimentally measured data and FLUKA simulated results has been observed.

Pu(V) transport through Savannah River Site soils - an evaluation of a conceptual model of surface- mediated reduction to Pu (IV)

Over the last fifteen years the Savannah River Site (SRS) in South Carolina, USA, was selected as the site of three new plutonium facilities: the Mixed Oxide Fuel Fabrication Facility, Pit Disassembly and Conversion Facility, and the Pu Immobilization Plant. In order to assess the potential human and environmental risk associated with these recent initiatives, improved understanding of the fate and transport of Pu in the SRS subsurface environment is necessary. The hypothesis of this study was that the more mobile forms of Pu, Pu(V) and Pu(VI), would be reduced to the less mobile Pu(III/IV) oxidation states under ambient SRS subsurface conditions. Laboratory-scale dynamic flow experiments (i.e., column studies) indicated that Pu(V) was very mobile in SRS sediments. At higher pH values the mobility of Pu decreased and the fraction of Pu that became irreversibly sorbed to the sediment increased, albeit, only slightly. Conversely, these column experiments showed that Pu(IV) was essentially immobile and was largely irreversibly sorbed to the sediment. More than 100 batch sorption experiments were also conducted with four end-member sediments, i.e., sediments that include the chemical, textural, and mineralogical properties likely to exist in the SRS. These tests were conducted as a function of initial Pu oxidation state, pH, and contact time and consistently demonstrated that although Pu(V) sorbed initially quite weakly to sediments, it slowly, over the course of <33 days, sorbed very strongly to sediments, to approximately the same degree as Pu(IV). This is consistent with our hypothesis that Pu(V) is reduced to the more strongly sorbing form of Pu, Pu(IV). These studies provide important experimental support for a conceptual geochemical model for dissolved Pu in a highly weathered subsurface environment. That is that, irrespective of the initial oxidation state of the dissolved Pu introduced into a SRS sediment system, Pu(IV) controls the environmental transport within a couple weeks and Pu strongly binds to the sediment, limiting its mobility.

Prediction analysis of dose equivalent responses of neutron dosemeters used at a MOX fuel facility

To predict how accurately neutron dosemeters can measure the neutron dose equivalent (rate) in MOX fuel fabrication facility work environments, the dose equivalent responses of neutron dosemeters were calculated by the spectral folding method. The dosemeters selected included two types of personal dosemeter, namely a thermoluminescent albedo neutron dosemeter and an electronic neutron dosemeter, three moderator-based neutron survey meters, and one special instrument called an H(p)(10) monitor. The calculations revealed the energy dependences of the responses expected within the entire range of neutron spectral variations observed in neutron fields at workplaces.

ＭＯＸ燃料加工施設における臨界事象発生頻度概略評価手法の開発

A criticality accident in a MOX fuel fabrication facility may occur depending on several parameters, such as mass inventory and plutonium enrichment. MOX handling units in the facility are designed and operated based on the double contingency principle to prevent criticality accidents. Control failures of at least two parameters are needed for the occurrence of criticality accident. To evaluate the probability of such control failures, the criticality conditions of each parameter for a specific handling unit are necessary for accident scenario analysis to be clarified quantitatively with a criticality analysis computer code. In addition to this issue, a computer-based control system for mass inventory is planned to be installed into MOX handling equipment in a commercial MOX fuel fabrication plant. The reliability analysis is another important issue in evaluating the likelihood of control failure caused by software malfunction. A likelihood estimation method for criticality accident has been developed with these issues been taken into consideration. In this paper, an example of analysis with the proposed method and the applicability of the method are also shown through a trial application to a model MOX fabrication facility.

Plutonium oxide polishing for MOX fuel fabrication

Los Alamos National Laboratory (LANL) successfully polished 120 kg of plutonium from surplus nuclear weapons for the European Mixed Oxide (MOX) Lead Test Assembly (LTA), managed by Duke, COGEMA, and Stone & Webster (DCS). The purified oxide was fabricated into fuel pellets, which comprised the LTAs. The LTAs have been inserted into the Catawba (SC, USA) nuclear reactor, where they are presently being used to generate electricity. The material used in the MOX LTA represents the first plutonium oxide from LANL to be polished under the quality requirements set forth by the Nuclear Regulatory Commission (NRC). In addition, the MOX project has been hailed as the “largest single nonproliferation project in history” by US Ambassador Linton Brooks (23/9/04). The process flow for PuO 2 purification was based on aqueous recovery that included various unit operations (dissolution, ion exchange, oxalate precipitation, and calcination). Data from a variety of chemical and physical analyses demonstrate product quality and process consistency. Further process optimization experiments will be performed prior to polishing an additional 330 kg of PuO 2 for the Mixed Oxide Fuel Fabrication Facility (MFFF) at Savannah River Site. During LTA production in 2004, approximately 500 L of 7 M nitric acid was used for washing impurities from the plutonium in order to meet product purity specifications. The majority of this acid was sent to the LANL Radioactive Liquid Waste Treatment Facility (TA-50) as effluent. Optimization experiments during 2006 and 2007 will investigate the use of recycled HNO 3 for PuO 2 washing and ion exchange. If successful, aqueous polishing processes will re-use almost all of the nitric acid that would otherwise have been sent to TA-50. In addition, an overall reduction in the volume of nitric acid used during ion exchange will be tested. The use of new Inconel calcination boats will also be verified. Additional process optimization activities will be conducted to validate the use of 100% quaternized Reillex ion exchange resin. Qualification exercises will include a series of process runs, followed by chemical analyses to assess product oxide impurity levels. Once the optimized processes have been qualified, the production phase of the project will begin.

Investigation of a Model to Evaluate the Pyrolysis Properties of Zinc Stearate

In the preparation process of a MOX powder for a MOX fuel fabrication facility, zinc stearate will be added into the MOX powder as an additional material to adjust the density. If too much of the material is added by mistake, the criticality characteristics of the MOX fuel will be influenced because of its neutron moderation effect. If a criticality condition should be induced by the excess zinc stearate, melting and pyrolysis of the material could be caused by the fission energy, and, therefore, the dynamic characteristics of the MOX fuel would be affected in a feedback mechanism. To provide a quantitative evaluation of such dynamics, thermal properties such as exothermic and endothermic calorific values, reaction rates, etc. were measured and correlated with the respective physical variations and release behavior of pyrolysis gas with the pyrolysis reaction. It was found that the exo/endothermic behavior of the material with rising temperature could be divided into six regions and that a rapid rise in pressure occurred over about 400°C caused by the pyrolysis reaction. Furthermore, on the basis of these results, an evaluation model for the thermal properties under criticality conditions was also investigated.

MOX燃料加工施設に対する確率論的安全評価手法の開発

A Probabilistic Safety Assessment (PSA) procedure for Mixed Oxide (MOX) fuel fabrication facilities was developed. The procedure is a "two-part five-step" approach which takes characteristics of MOX fuel fabrication facilities into consideration. In the first part, so-called preliminary PSA, the hazard analysis approach was applied, which consists of two analysis steps: Functional Failure Modes and Effects Analysis (FFMEA) and Risk Matrix Analysis. The FFMEA analyzes a variety of functions of equipment composing the facility to identify potential abnormal events exhaustively. In the Risk Matrix Analysis, these potential events are screened to select abnormal events as candidates to be analyzed in the second part, using two-dimensional matrix based on the likelihood evaluated by probabilistic index method and maximum unmitigated radioactive release calculated by the Five-Factor Formula. For the selected abnormal events, in the second part, so-called detailed PSA, accident sequences, their occurrence frequencies and consequences are analyzed. These three analysis steps correspond to PSA procedure for nuclear power plant. The applicability of the PSA procedure was demonstrated through the trial application to model plant of MOX fuel fabrication facility.

Use of Sensitivity and Uncertainty Analysis to Select Benchmark Experiments for the Validation of Computer Codes and Data

Sensitivity and uncertainty analysis methodologies under development at Oak Ridge National Laboratory were applied to determine whether existing benchmark experiments adequately cover the area of applicability for the criticality code and data validation of PuO2 and mixed-oxide (MOX) powder systems. The study examined three PuO2 powder systems and four MOX powder systems that would be useful for establishing mass limits for a MOX fuel fabrication facility. Using traditional methods to choose experiments for criticality analysis validation, 46 benchmark critical experiments were identified as applicable to the PuO2 powder systems. However, only 14 experiments were thought to be within the area of applicability for dry MOX powder systems.The applicability of 318 benchmark critical experiments, including the 60 experiments initially identified, was assessed. Each benchmark and powder system was analyzed using the Tools for Sensitivity and UNcertainty Analysis Methodology Implementation (TSUNAMI) one-dimensional (TSUNAMI-1D) or TSUNAMI three-dimensional (TSUNAMI-3D) sensitivity analysis sequences, which will be included in the next release of the SCALE code system. This sensitivity data and cross-section uncertainty data were then processed with TSUNAMI-IP to determine the correlation of each application to each experiment in the benchmarking set. Correlation coefficients are used to assess the similarity between systems and determine the applicability of one system for the code and data validation of another.The applicability of most of the experiments identified using traditional methods was confirmed by the TSUNAMI analysis. In addition, some PuO2 and MOX powder systems were determined to be within the area of applicability of several other benchmarks that would not have been considered using traditional methods. Therefore, the number of benchmark experiments useful for the validation of these systems exceeds the number previously expected. The TSUNAMI analysis also emphasized some areas where more benchmark data are needed, indicating the need for further evaluation of existing experiments, or possibly the completion of new experiments to fill these gaps. This lack of evaluated data is particularly important for very dry and dense MOX powder systems.

Plutonium aerosol characterization inside safety enclosures at a demonstration mixed-oxide fuel fabrication facility.

Plutonium aerosol characterization inside safety enclosures at a demonstration mixed-oxide fuel fabrication facility.

Ambulatory Pneumothorax in Pulmonary Tuberculosis

Los Alamos National Laboratory (LANL) successfully polished 120 kg of plutonium from surplus nuclear weapons for the European Mixed Oxide (MOX) Lead Test Assembly (LTA), managed by Duke, COGEMA, and Stone & Webster (DCS). The purified oxide was fabricated into fuel pellets, which comprised the LTAs. The LTAs have been inserted into the Catawba (SC, USA) nuclear reactor, where they are presently being used to generate electricity. The material used in the MOX LTA represents the first plutonium oxide from LANL to be polished under the quality requirements set forth by the Nuclear Regulatory Commission (NRC). In addition, the MOX project has been hailed as the “largest single nonproliferation project in history” by US Ambassador Linton Brooks (23/9/04). The process flow for PuO2 purification was based on aqueous recovery that included various unit operations (dissolution, ion exchange, oxalate precipitation, and calcination). Data from a variety of chemical and physical analyses demonstrate product quality and process consistency. Further process optimization experiments will be performed prior to polishing an additional 330 kg of PuO2 for the Mixed Oxide Fuel Fabrication Facility (MFFF) at Savannah River Site. During LTA production in 2004, approximately 500 L of 7 M nitric acid was used for washing impurities from the plutonium in order to meet product purity specifications. The majority of this acid was sent to the LANL Radioactive Liquid Waste Treatment Facility (TA-50) as effluent. Optimization experiments during 2006 and 2007 will investigate the use of recycled HNO3 for PuO2 washing and ion exchange. If successful, aqueous polishing processes will re-use almost all of the nitric acid that would otherwise have been sent to TA-50. In addition, an overall reduction in the volume of nitric acid used during ion exchange will be tested. The use of new Inconel calcination boats will also be verified. Additional process optimization activities will be conducted to validate the use of 100% quaternized Reillex ion exchange resin. Qualification exercises will include a series of process runs, followed by chemical analyses to assess product oxide impurity levels. Once the optimized processes have been qualified, the production phase of the project will begin.