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All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Decommissioning, Dismantling and Disarming: a Unique Information Showroom Inside the G2 Reactor at Marcoule Centre (France) - 12068

The paper aims at presenting the new information showroom called 'Escom G2' (for 'Espace Communication') inaugurated by the French Atomic Energy and Alternative Energies Commission (CEA) in spring 2011. This showroom is settled directly inside the main building of the G2 nuclear reactor: a facility formerly dedicated to weapon-grade plutonium production since the late 1950's at the Marcoule nuclear centre, in south of France. After its shutdown, and reprocessing of the last spent fuels, a first dismantling step was successfully completed from 1986 to 1996. Unique in France and in Europe, Escom G2 is focused on France dismantling expertise and its action for disarmament. This showroom comprises of a 300-square meters permanent exhibition, organized around four themes: France strategy for disarmament, decommissioning and dismantling technical aspects, uranium and plutonium production cycles. Each of these topics is illustrated with posters, photos, models and technical pieces from the dismantled plants. It is now used to present France's action in disarmament to highly ranked audiences such as: state representatives, diplomats, journalists... The paper explains the background story of this original project. As a matter of fact, in 1996 France was the first nuclear state to decide to shut down and dismantle its fissile material production facilities for nuclear weapons. First, the paper presents the history of the G2 reactor in the early ages of Marcoule site, its operating highlights as well as its main dismantling operations, are presented. In Marcoule, where the three industrial-scale reactors G1, G2 and G3 used to be operated for plutonium production (to be then reprocessed in the nearby UP1 plant), the initial dismantling phase has now been completed (in 1980's for G1 and in 1996 for G2 and G3). The second phase, aimed at completely dismantling these three reactors, will restart in 2020, and is directly linked to the opening of a future national storage facility for irradiated graphite waste. Then, the paper recalls communication events and official visits hosted in Pierrelatte and Marcoule, following a formal invitation from the French President Mr. Nicolas Sarkozy. These visits, which were organized in order to illustrate the irreversibility of these dismantling operations, allowed visitors to discovers places that used to be former highly classified areas. Three official visits were organized in 2008 and 2009 for representatives of the Conference on Disarmament Member States, non-governmental experts and journalists. All participants visited the dismantled uranium enrichment plant in Pierrelatte, the G2 reactor and the UP1 plant in Marcoule. The visits were successful and visitors were especially impressed by the G2 reactor and its massive industrial architecture, symbolic of the early ages of nuclear history. In late 2010, this feedback convinced CEA Military Application Directorate (CEA DAM) that a permanent showroom could be installed inside the reactor, making it possible to preserve the cultural value of this historical landmark, and to continue its ongoing effort of communication and outreach. The paper explains the design of this concept: the museography project with a professional designer, the communication material conception and the features of such an original place. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Study of the influence of heat sources on the out-of-pile calibration curve of calorimetric cells used for nuclear energy deposition quantification

At present the Jules Horowitz Reactor is under construction in Cadarache research center of CEA 'French Alternative Energies and Atomic Energy Commission' center located in the south-east of France. This new Material Testing Reactor (MTR) will be operational in late 2019 and will allow the generation of a new experimental potential (up to 20 irradiation devices simultaneously) and new harsh conditions such as higher neutron fluxes (5.10{sup 14} n.cm{sup -2}.s{sup -1} for E≥1 MeV), faster material ageing and higher nuclear heating (up to 20 W/g for nominal capacity of 100 MW). In nuclear research field, the control and the measurement of the nuclear heating (energy deposition rate per mass unit induced by the interactions of radiations with matter) is crucial to carry out accurate studies on ageing of materials and on the behavior of nuclear fuels under irradiation. Several experiments need to know precisely this key parameter in order to establish dedicated thermal conditions. The measurement of the nuclear heating inside MTRs is realized by three kinds of sensors: single-cell calorimeter, differential calorimeter and gamma thermometer. One scientific objective of the IN-CORE program, between CEA and Aix-Marseille University in 2009, is to improve the nuclear heating measurement. In this context a new multi-sensor device, called CARMEN, was made. This device contains in particular a differential calorimeter which was designed to measure the nuclear heating in the periphery of OSIRIS reactor (a MTR located at Saclay, France) up to 2 W/g and tested during two irradiation campaigns. Results obtained during these campaigns showed that temperatures reached inside the calorimeter are higher than ones obtained during the preliminary out-of-pile calibration experiments. For instance for 1.74 W/g, the in-pile temperature of the calorimeter rod is equal to 305 deg. C against 225 deg. C in laboratory conditions by simulating the nuclear heating by Joule Effect inside the calorimeter cell head. This discrepancy is higher than in previous experiments because the calorimeter owns a high sensitivity. Consequently, a new prototype was created and instrumented by other heat sources in order to impose an energy deposition on the calorimetric cell structure (in particular in the base) and to improve the calibration step in out-of-pile conditions. In this paper, on the first part a detailed description of the new calorimetric sensor will be given. On the second part, the experimental response of the sensor obtained for several internal heating conditions will be shown. The influence of these conditions on the calibration curve will be discussed. Then the response of this prototype will be also presented for different external cooling fluid conditions (in particular flow temperature). In this part, the comparison between the in-pile and out-of-pile experimental results will be performed. On the last part, these out-of-pile experiments will be completed by 2D axisymmetrical thermal simulations with the CEA code CAST3M using Finite Elements Method. After a comparison between experimental and numerical works, improvements of the sensor prototype will be studied (new heat sources). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

French Atomic and alternative energies commission decommissioning policy and strategies: What is at stake lessons learned and feedback experience - 15013

Since its creation in 1945, the French Atomic Energy and Alternative Energies Commission has carried out research programmes on civil use of nuclear, and its application. CEA has set up and run various types of installations from research (accelerators, pond reactors, fast reactors, laboratories) to industry (power reactor, processing facilities). There are currently more than several tens of facilities which are presently closed down pending dismantling or in the process of being dismantled by CEA, the governmental body for research and development in nuclear and renewable energies. When operating, all these installations have generated radioactive waste. When they are shutdown, either as a result of the programmes for which they were set up, or due to equipment wearing out or being obsolete, these installations must be cleaned up, dismantled and decommissioned, taking into account their nuclear nature and requirements related to safety and environment. During the decommissioning process, the implementer has to deal with safety and site remediation issues. Dealing with these installations is a long-term programme which must be carefully planned and financed in order to optimise its management. The paper will describe the CEA's decommissioning policy, strategies adopted, and draw up lessons learned and feedback experience, showing that: - Safety is guaranteed, - Dismantling is a technically controllable operation, but R and D actions are to be developed for specific problems as characterization or operation in hostile environment. Moreover all the methods and techniques improve as a result of feedback experience, - The amount of material and waste produced by the dismantling are manageable. The very low, low and intermediate level waste can be stored on existing shallow land burial centres, some new repositories have to be created for specific waste (graphite, radium,...) and for high and intermediate level long live waste, which are stored in interim storage , - The dismantling must be optimized, especially by appliance of a good strategy based on rational industrial choices. - Communication and transparency with stakeholders is essential for global process acceptance, - The lessons learned and how feedback experience could benefit for future decommissioning projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Benefits from R. and D. for D. and D. projects - 15089

CEA (French Alternative Energies and Atomic Energy Commission) is both the operator of important nuclear facilities all over the nuclear cycle, in charge of major new built or D and D projects and a R and D group with dynamic policy of technology transfer: second in France for patents deposit, active partnership agreements with industries, and support for Start-ups. The position of CEA in D and D is unique because of the number and the wide diversity of facilities under decommissioning, with some high level of contamination. Innovative solutions are being developed in 6 main axes to protect the operators, to minimize the overall costs and the volumes of waste: - Investigations in the facilities - Radiological measurement of waste - Robotics, laser cutting devices, virtual reality to simulate and choose the best way to operate, - Decontamination of soils and structures - Waste treatment and conditioning - Methods and Information Technology (IT)Tools for project and waste management The last developments are shown and examples of industrial applications given: encapsulation of Magnesium or vitrification of high level waste for UP1 retreatment plant, treatment of sodium for Phenix fast breeder, treatment of organic liquids, etc. CEA is willing to share actions in partnership with other operators or with industrials so they can use the results for other contracting authorities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Applicability Evaluation of the In-can Vitrification to Fukushima Waste - 19016

The in-can vitrification process has been developed by the CEA for an internal application (alpha waste treatment) since 2005 and can be rated TRL 7. On these basis the DEM'n'MELT consortium which gathers Orano, CEA and ECM technologies is currently developing and qualifying a robust, simple and versatile in situ vitrification process dedicated to D and D waste. The DEM'n'MELT vitrification process is developed with a modular design, the main characteristics are presented hereafter: - Suitable for high or medium level waste. - Heated by a simple and robust resistance furnace. - Offering homogeneity to the melted mixture. - Suitable for solid or liquid wastes, including zeolites and sludge. - Flexible enough to accommodate uncertainties on waste composition. - Designed in order to produce a small amount of secondary waste. - Implementing proven technological solutions from French high level vitrification shops. - Requiring low investment and operation costs. Regarding this features, the in-can process has been deemed as an appropriate tool for the vitrification of the Fukushima water treatment secondary waste (mineral adsorbants, sludge). The evaluation of the In-can process for Fukushima waste is a project led by Orano, CEA and ANADEC. The main phases are: - Scenarios and glass formulation studies to optimize waste loading and matrix homogeneity. - Bench scale tests on several scenarios of blended waste. - Pilot scale test on one selected scenario (reference scenario). - Feasibility evaluation to Fukushima Daiichi Nuclear Power Station implementation. This paper presents the first results of this project with a highlighting on glass waste loading, glass properties, bench tests and pilot test results, process parameters, and material balance. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Comparisons between a priori Uncertainty Quantification and Calculation/Measurement Discrepancies Applied to the MERCI UO{sub 2} Fuel Rod Decay Heat Experiment

In 2008, an UO{sub 2} fuel rod was irradiated up to ∼ 3.5 GWd/tHL in the CEA/Saclay research reactor Osiris through the MERCI experiment. Some rod's pellets were analysed. Experimental results were obtained: - fuel rod decay heat measurement by calorimetry (cooling times from 27 minutes to 42 days), - evaluation of the amounts of some nuclei (U, Pu, Cs, Nd...) by isotopic dilution mass spectrometry, - nuclide activities. Comparison of type (C - M)/M where C is the calculated decay heat and M the experimentally measured decay heat was reported in reference. Neutronics flux was computed by CEA TRIPOLI-4 Monte-Carlo or APOLLO2 deterministic transport code system. Irradiation and cooling phases were simulated with CEA DARWIN/PEPIN2 code system. Good agreements were globally shown between numerical simulation results and experimental measurements on total decay heat. Uncertainty propagation in this previous work was established from Rebah's studies. These last years, within the framework of supports to its industrial partners EDF and Areva, the CEA/DEN has implemented two different approaches to propagate nuclear data uncertainties to depletion code outputs. A direct forward deterministic perturbation method using sensitivity profiles was implemented in the French industrial depletion code system DARWIN/PEPIN2. Meanwhile, a propagation based on Monte Carlo correlated sampling was achieved using the CEA uncertainty platform URANIE and the new generation depletion code MENDEL. Both methods will be described in this article. Uncertainties on nuclear data (independent fission yields, multigroup microscopic cross sections, radioactive decay constants, radioactive decay branching ratio and radioactive decay energies) are propagated to decay heat and isotopic concentrations. To match MERCI first study, flux computation is realized by APOLLO2. Furthermore, uncertainty quantification (UQ) is re-evaluated by both DARWIN/PEPIN2 and MENDEL codes. It is an important stage in the process of validating both depletion codes with their respective data library. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Strong‐Yet‐Ductile Eutectic Alloys Employing Cocoon‐Like Nanometer‐Sized Dislocation Cells

Abstract Eutectic alloys (EAs) with superior fluidity are known to be the easiest to cast into high‐quality ingots, making them the alloys of choice for making large‐sized structural parts. However, conventional EAs (CEAs) have never reached strength–ductility combinations on par with the best in other alloy categories. Via thermomechanical processing of cast Ni‐32.88wt%Fe‐9.53wt%Al CEAs, a cocoon‐like nano‐meshed (as fine as 26 nm) network of dislocations (CNN‐D) is produced via recovery annealing, through the rearrangement of cold‐work‐accumulated dislocations anchored by dense pre‐existing nanoprecipitates. In lieu of traditional plasticity mechanisms, such as TWIP and TRIP, the CNN‐D is particularly effective in eutectic lamellae with alternating phases, as it instigates nanometer‐spaced planar slip bands that not only dynamically refine the microstructure but also transmit from the FCC (face‐centered‐cubic) layers into the otherwise brittle B2 layers. These additional mechanisms for strengthening and strain hardening sustain stable tensile flow, resulting in a striking elevation of both strength and ductility to outrank not only all previous CEAs, but also the state of the art—additively manufactured eutectic high‐entropy alloys. The CNN‐D thus adds a novel microstructural strategy for performance enhancement, especially for compositionally complex alloys that increasingly make use of nanoprecipitates or local chemical order.

Chemistry↗
DOE OSTI2011

Alchemy to reason: Effective use of Cumulative Effects Assessment in resource management

Cumulative Effects Assessment (CEA) is a tool that can be useful in making decisions about natural resource management and allocation. The decisions to be made include those (i) necessary to construct planning and regulatory frameworks to control development activity so that societal goals will be achieved and (ii) whether or not to approve individual development projects, with or without conditions. The evolution of CEA into a more successful tool cannot occur independently of the evolution of decision making processes. Currently progress is painfully slow on both fronts. This paper explores some opportunities to accelerate improvements in decision making in natural resource management and in the utility of CEA as a tool to assist in making such decisions. The focus of the paper is on how to define the public interest by determining what is acceptable.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2011

Development and Validation of a Prognostic Model Using Blood Biomarker Information for Prediction of Survival of Non-Small-Cell Lung Cancer Patients Treated With Combined Chemotherapy and Radiation or Radiotherapy Alone (NCT00181519, NCT00573040, and NCT00572325)

Purpose: Currently, prediction of survival for non-small-cell lung cancer patients treated with (chemo)radiotherapy is mainly based on clinical factors. The hypothesis of this prospective study was that blood biomarkers related to hypoxia, inflammation, and tumor load would have an added prognostic value for predicting survival. Methods and Materials: Clinical data and blood samples were collected prospectively (NCT00181519, NCT00573040, and NCT00572325) from 106 inoperable non-small-cell lung cancer patients (Stages I-IIIB), treated with curative intent with radiotherapy alone or combined with chemotherapy. Blood biomarkers, including lactate dehydrogenase, C-reactive protein, osteopontin, carbonic anhydrase IX, interleukin (IL) 6, IL-8, carcinoembryonic antigen (CEA), and cytokeratin fragment 21-1, were measured. A multivariate model, built on a large patient population (N = 322) and externally validated, was used as a baseline model. An extended model was created by selecting additional biomarkers. The model's performance was expressed as the area under the curve (AUC) of the receiver operating characteristic and assessed by use of leave-one-out cross validation as well as a validation cohort (n = 52). Results: The baseline model consisted of gender, World Health Organization performance status, forced expiratory volume, number of positive lymph node stations, and gross tumor volume and yielded an AUC of 0.72. The extended model included two additional blood biomarkers (CEA and IL-6) and resulted in a leave-one-out AUC of 0.81. The performance of the extended model was significantly better than the clinical model (p = 0.004). The AUC on the validation cohort was 0.66 and 0.76, respectively. Conclusions: The performance of the prognostic model for survival improved markedly by adding two blood biomarkers: CEA and IL-6.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2011

Irradiation tests of prototype self-powered gamma and neutron detectors

In the framework of the SCK.CEN-CEA Joint Instrumentation Laboratory, we are developing and optimizing a self-powered detector for selective in-core monitoring of the gamma field. Several prototypes with bismuth emitters were developed and tested in a pure gamma field (the PAGURE gamma irradiation facility at CEA) and in mixed neutron and gamma fields (in the OSIRIS reactor at CEA and in the BR2 reactor at SCK.CEN). Detailed MCNP modelling was performed to calculate the gamma and neutron sensitivities. Apart from a few failing prototypes, all detectors showed equilibrium signals proportional to the gamma field with a good long-term stability (under irradiation during several weeks). A tubular geometry design was finally selected as the most appropriate for in-core gamma detection, coupling a larger sensitivity with better response characteristics. In the same experiment in BR2 six prototype Self-Powered Neutron Detectors (SPNDs) with continuous sheaths (i.e. without any weld between the sensitive part and the cable) were extensively tested: two SPNDs with Co emitter, two with V emitter and two with Rh emitters, with varying geometries. All detector responses were verified to be proportional to the reactor power. The prompt and delayed response contributions were quantified. The signal contributions due to the impact of gamma rays were experimentally determined. The evolution of the signals was continuously followed during the full irradiation period. The signal-to-noise level was observed to be well below 1% in typical irradiation conditions. The absolute neutron and gamma responses for all SPNDs are consistent. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Materials Data on CeAsN(O2F)2 by Materials Project

(CeAs(O2F)2)2N2 crystallizes in the monoclinic P2_1/m space group. The structure is three-dimensional and consists of two ammonia molecules and one CeAs(O2F)2 framework. In the CeAs(O2F)2 framework, Ce4+ is bonded in a 8-coordinate geometry to four O2- and four equivalent F1- atoms. There are a spread of Ce–O bond distances ranging from 2.24–2.55 Å. There are two shorter (2.31 Å) and two longer (2.32 Å) Ce–F bond lengths. As5+ is bonded in a tetrahedral geometry to four O2- atoms. There is one shorter (1.71 Å) and three longer (1.72 Å) As–O bond length. There are three inequivalent O2- sites. In the first O2- site, O2- is bonded in a distorted single-bond geometry to one Ce4+ and one As5+ atom. In the second O2- site, O2- is bonded in a bent 150 degrees geometry to one Ce4+ and one As5+ atom. In the third O2- site, O2- is bonded in a distorted bent 150 degrees geometry to one Ce4+ and one As5+ atom. F1- is bonded in a distorted bent 120 degrees geometry to two equivalent Ce4+ atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2022

PIP-II 650 MHz Power Coupler Thermal Studies

The Proton Improvement Plan - II (PIP-II) project is underway at Fermilab with an international collaboration involving CEA in the development and testing of 650 MHz cryomodules. One of the first main contributions of the CEA was the participation in the design efforts for the current 50 KW CW 650 MHz power couplers. This paper reports some of the results of thermal and paramet-ric studies carried out by the CEA on these power couplers

43 PARTICLE ACCELERATORS↗
DOE OSTI2013

ATRC Neutron Detector Testing Quick Look Report

As part of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program, a joint Idaho State University (ISU) / French Alternative Energies and Atomic Energy Commission (CEA) / Idaho National Laboratory (INL) project was initiated in FY-10 to investigate the feasibility of using neutron sensors to provide online measurements of the neutron flux and fission reaction rate in the ATR Critical Facility (ATRC). A second objective was to provide initial neutron spectrum and flux distribution information for physics modeling and code validation using neutron activation based techniques in ATRC as well as ATR during depressurized operations. Detailed activation spectrometry measurements were made in the flux traps and in selected fuel elements, along with standard fission rate distribution measurements at selected core locations. These measurements provide additional calibration data for the real-time sensors of interest as well as provide benchmark neutronics data that will be useful for the ATR Life Extension Program (LEP) Computational Methods and V&V Upgrade project. As part of this effort, techniques developed by Prof. George Imel will be applied by Idaho State University (ISU) for assessing the performance of various flux detectors to develop detailed procedures for initial and follow-on calibrations of these sensors. In addition to comparing data obtained from each type of detector, calculations will be performed to assess the performance of and reduce uncertainties in flux detection sensors and compare data obtained from these sensors with existing integral methods employed at the ATRC. The neutron detectors required for this project were provided to team participants at no cost. Activation detectors (foils and wires) from an existing, well-characterized INL inventory were employed. Furthermore, as part of an on-going ATR NSUF international cooperation, the CEA sent INL three miniature fission chambers (one for detecting fast flux and two for detecting thermal flux) with associated electronics for assessment. In addition, Prof. Imel, ISU, has access to an inventory of Self-Powered Neutron Detectors (SPNDs) with a range of response times as well as Back-to-Back (BTB) fission chambers from prior research he conducted at the Transient REActor Test Facility (TREAT) facility and Neutron RADiography (NRAD) reactors. Finally, SPNDs from the National Atomic Energy Commission of Argentina (CNEA) were provided in connection with the INL effort to upgrade ATR computational methods and V&V protocols that are underway as part of the ATR LEP. Work during fiscal year 2010 (FY10) focussed on design and construction of Experiment Guide Tubes (EGTs) for positioning the flux detectors in the ATRC N-16 locations as well as obtaining ATRC staff concurrence for the detector evaluations. Initial evaluations with CEA researchers were also started in FY10 but were cut short due to reactor reliability issues. Reactor availability issues caused experimental work to be delayed during FY11/12. In FY13, work resumed; and evaluations were completed. The objective of this "Quick Look" report is to summarize experimental activities performed from April 4, 2013 through May 16, 2013.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

LLNL Results from CALIBAN-PROSPERO Nuclear Accident Dosimetry Experiments in September 2014

Lawrence Livermore National Laboratory (LLNL) uses thin neutron activation foils, sulfur, and threshold energy shielding to determine neutron component doses and the total dose from neutrons in the event of a nuclear criticality accident. The dosimeter also uses a DOELAP accredited Panasonic UD-810 (Panasonic Industrial Devices Sales Company of America, 2 Riverfront Plaza, Newark, NJ 07102, U.S.A.) thermoluminescent dosimetery system (TLD) for determining the gamma component of the total dose. LLNL has participated in three international intercomparisons of nuclear accident dosimeters. In October 2009, LLNL participated in an exercise at the French Commissariat à l’énergie atomique et aux énergies alternatives (Alternative Energies and Atomic Energy Commission- CEA) Research Center at Valduc utilizing the SILENE reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison at CEA Valduc, this time with exposures at the CALIBAN reactor (Hickman et al. 2011). This paper discusses LLNL’s results of a third intercomparison hosted by the French Institut de Radioprotection et de Sûreté Nucléaire (Institute for Radiation Protection and Nuclear Safety- IRSN) with exposures at two CEA Valduc reactors (CALIBAN and PROSPERO) in September 2014. Comparison results between the three participating facilities is presented elsewhere (Chevallier 2015; Duluc 2015).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report

A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chamber and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A project meeting was held at KSU to discuss the roles and responsibilities between INL and KSU for development of the HT MPFDs. Provide input to various irradiation programs for installation of the MPFD technology in irradiation tests. As documented in this report, FY15 funding has allowed the project to meet year one planned accomplishments to develop a HT MPFD that offers US MTR users enhanced capabilities for real-time measurement of flux and temperature with a single detector. In addition, the accomplishments of this project have attracted funding from other Department of Energy Office of Nuclear Energy (DOE-NE) programs for additional applications. The work in those programs will build on current activities completed in this NEETASI HT MPFD project, but the MPFD will be specifically tailored to meet their program needs.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Baseline Postirradiation Examination of the FUTURIX-FTA Experiments

Postirradiation Examination (PIE) results from the baseline examination of the FUTURIX-FTA irradiation experiment are presented in this report. The FUTURIX-FTA irradiation was designed to study the fuel performance of candidate transmutation fuels. The FUTURIX-FTA experiment was conducted in the true fast neutron spectrum conditions of the Phénix sodium fast test reactor in France. FUTURIX-FTA contained several compositions that were also irradiated as part of the AFC-1 test in cadmium filtered positions in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) that approximate the neutron spectrum of a fast reactor. This experiment was a joint collaboration between the Department of Energy (DOE) in the U.S. and the Commissariat á l'Energie Atomique et aux Energies Alternatives (CEA) in France. A joint agreement between DOE and CEA was established in 2004 to conduct an irradiation experiment of metallic and nitride fuels in the Phénix reactor at CEA. Both the FUTURIX-FTA tests and the AFC-1 tests were conducted for the DOE Advanced Fuels Campaign (AFC) that is part of Fuel Cycle Research & Development (FCRD) program. These programs seek to develop and demonstrate the technologies needed to transmute long-lived transuranic actinide isotopes contained in spent nuclear fuel via fast reactor technology. The long term goal of this work is to reduce the radiotoxicity of future high level waste repositories. Post-irradiation examination of irradiated fuels experiments provides data related to in reactor fuel performance and input into future fuel design choices. The compositions of the FUTURIX-FTA pins were designed to test if minor actinides could be incorporated into the fuel of a fast reactor without significantly changing fuel performance observed for previously well studied fuels. The fuel performance of metallic fuel in fast reactors has been well documented. Likewise the irradiation performance of nitride fuels in fast reactors has also been documented. However the existing database for nitride fuels is much smaller than the database for metal fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Performance Benchmark of Commercial and Developmental Fission Chambers in Elevated Temperatures

This report documents the testing of two in-core fission chamber technologies for high-temperature irradiation environments. This work is in collaboration with the French Alternative Energies and Atomic Energy Commission (CEA). The fission chamber evaluated by Idaho National Laboratory is the micro-pocket fission detector (MPFD). The fission chamber evaluated by the CEA are the 3 mm miniaturized fission chamber and the 7 mm high-temperature fission chamber. Demonstrations of the MPFD were performed at the Neutron Radiography Facility and the Massachusetts Institute of Technology Reactor. Demonstrations of the CEA fission chambers were performed at the Ohio State University Research Reactor. All demonstrations were conducted with a heated experiment rig up to 850°C.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Characterization and Valuation of the Uncertainty of Calibrated Parameters in Microsimulation Decision Models

We evaluated the implications of different approaches to characterize the uncertainty of calibrated parameters of microsimulation decision models (DMs) and quantified the value of such uncertainty in decision making. We calibrated the natural history model of CRC to simulated epidemiological data with different degrees of uncertainty and obtained the joint posterior distribution of the parameters using a Bayesian approach. We conducted a probabilistic sensitivity analysis (PSA) on all the model parameters with different characterizations of the uncertainty of the calibrated parameters. We estimated the value of uncertainty of the various characterizations with a value of information analysis. We conducted all analyses using high-performance computing resources running the Extreme-scale Model Exploration with Swift (EMEWS) framework. The posterior distribution had a high correlation among some parameters. The parameters of the Weibull hazard function for the age of onset of adenomas had the highest posterior correlation of -0.958. When comparing full posterior distributions and the maximum-a-posteriori estimate of the calibrated parameters, there is little difference in the spread of the distribution of the CEA outcomes with a similar expected value of perfect information (EVPI) of $\$$653 and $\$$685, respectively, at a willingness-to-pay (WTP) threshold of $\$$66,000 per quality-adjusted life year (QALY). Ignoring correlation on the calibrated parameters’ posterior distribution produced the broadest distribution of CEA outcomes and the highest EVPI of $\$$809 at the same WTP threshold. Different characterizations of the uncertainty of calibrated parameters affect the expected value of eliminating parametric uncertainty on the CEA. Ignoring inherent correlation among calibrated parameters on a PSA overestimates the value of uncertainty.

97 MATHEMATICS AND COMPUTING↗