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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Dounreay Site's Interim End State-Environmental Safety Strategy - 17509

The Dounreay nuclear licensed site, located in Caithness on the north coast of Scotland, was the United Kingdom's (UK) site for Fast Reactor research and development. The Dounreay facilities are now being decommissioned and the site restored in accordance with UK Government policy to manage civil nuclear liabilities. This site closure and environmental restoration programme is to achieve levels of residual radiological and chemical contamination sufficiently low that the site will be safe for future use. This document sets out the Environmental Safety Strategy for the site remediation and closure programme, which will follow completion of facility decommissioning and demolition. The document summarises the contractual definition of the Dounreay site end states and also the associated regulatory regimes. An overview of the approach and processes for achieving the Interim End State (IES) are provided, along with consideration of how Dounreay will demonstrate that the IES has been achieved. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Evaluation of Reduced Order Model for HT-9 Creep and Modifications to Current HT-9 Creep Model in BISON

In order to leverage existing reduced order models (ROM) for the modeling of LANL-developed HT-9 cladding constitutive behavior, mechanistic-based models have been implemented into BISON. It is posited that the increased fidelity of the ROM will show improvement over engineering-scale models in terms of comparison against experimental measurements. INL will assist LANL in implementing an HT9 ROM in BISON. Once the HT-9 mechanistic constitutive model can be leveraged in nuclear performance simulations, the numerical results of fast reactor models will be analyzed and compared against the Fuels Irradiation & Physics Database (FIPD) and separate effects tests to attempt a validation process. INL and LANL will work together to perform any necessary improvement that is identified during the implementation and use processes. Successful completion of the milestone will enable the general use of validated mechanistic ROMs for HT-9 cladding in BISON. This report provides a brief introduction to the Los Alamos Reduced Order Model Applied to Nonlinear Constitutive Equations (LAROMANCE) code, mechanical testing used to illustrate its creep and plastic deformation predictions for HT-9 cladding, and a comparison with existing models implemented into BISON. Current models have been modified to more appropriately account for primary thermal creep and results for corresponding testing are included in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

3S (Safeguards, Security, Safety) based pyroprocessing facility safety evaluation plan

The big advantage of pyroprocessing for the management of spent fuels against the conventional reprocessing technologies lies in its proliferation resistance since the pure plutonium cannot be separated from the spent fuel. The extracted materials can be directly used as metal fuel in a fast reactor, and pyroprocessing reduces drastically the volume and heat load of the spent fuel. KAERI has implemented the SBD (Safeguards-By-Design) concept in nuclear fuel cycle facilities. The goal of SBD is to integrate international safeguards into the entire facility design process since the very beginning of the design phase. This paper presents a safety evaluation plan using a conceptual design of a reference pyroprocessing facility, in which 3S (Safeguards, Security, Safety)-By-Design (3SBD) concept is integrated from early conceptual design phase. The purpose of this paper is to establish an advanced pyroprocessing hot cell facility design concept based on 3SBD for the successful realization of pyroprocessing technology with enhanced safety and proliferation resistance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Fuel cycle cost uncertainty from nuclear fuel cycle comparison

This paper examined the uncertainty in fuel cycle cost (FCC) calculation by considering both model and parameter uncertainty. Four different fuel cycle options were compared in the analysis including the once-through cycle (OT), the DUPIC cycle, the MOX cycle and a closed fuel cycle with fast reactors (FR). The model uncertainty was addressed by using three different FCC modeling approaches with and without the time value of money consideration. The relative ratios of FCC in comparison to OT did not change much by using different modeling approaches. This observation was consistent with the results of the sensitivity study for the discount rate. Two different sets of data with uncertainty range of unit costs were used to address the parameter uncertainty of the FCC calculation. The sensitivity study showed that the dominating contributor to the total variance of FCC is the uranium price. In general, the FCC of OT was found to be the lowest followed by FR, MOX, and DUPIC. But depending on the uranium price, the FR cycle was found to have lower FCC over OT. The reprocessing cost was also found to have a major impact on FCC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

International collaboration, the route to fuel cycle research and development

In hindsight, involvement with European Framework projects such as GoFastR (Gas-cooled Fast Reactors) and ACSEPT (Actinide Recycling by Separation and Transmutation) was a crucial and, at the time, an innovative step in maintaining the UK skills base during a period of major changes in the UK nuclear industry. It has undoubtedly delivered the objectives intended in terms of maintenance of the key skills, developing and training new staff, regenerating facilities and building strong links with the European nuclear research community. Over the last 2-3 years NNL's participation in European projects has moved forward such that NNL (National Nuclear Laboratory) is an integral partner of several major projects, fully engaged with delivering the core objectives of the projects and intent on forging deep collaborations with key organisations across Europe. With the renewed interest in nuclear energy and future fuel cycle options in the UK, NNL is now well positioned to contribute at an even deeper level in European level programmes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Multi-species Modeling of Molten-Salt Electrochemical Process

Spent nuclear fuel has been produced and stored in over 30 different countries which are operating commercial nuclear power plants. Molten-salt pyroprocessing has been developed for treating both oxide and metal spent nuclear fuels. This process separates long-lived actinide isotopes together as a group from fission products for burning the long-lived isotopes in fast reactors. Main processes of pyroprocessing such as oxide reduction, electrorefining, and electrowinning are based on a concept of electrolytic cell consuming energy to proceed electrochemical reactions. To scale-up the electrochemical processes, the design optimization of the key steps is required to reach high throughput that improves economic competitiveness and reduces an environmental impact. For the optimization, it is significant to compare different unit process designs and find optimum designs and operating conditions. However, this optimization process is very costly if we need to conduct experiments on every case with spent nuclear fuel. Instead, a validated simulation model can help to rapidly calculate and evaluate different case at significantly reduced costs. This paper presents recent development of multi-physics simulation of molten salt electrochemistry for partitioning spent nuclear fuel with limiting factors. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

A Variational Nodal Approach to 2D/1D Pin Resolved Neutron Transport: II Spherical Harmonics

The variational nodal method (VNM) as implemented in the VARIANT code has found wide application to fast reactor problems where fuel assembly homogenization does not result in unacceptable errors. The objective of this research is to generalize VNM for application to PWR problems, where homogenization even at the pin-cell level often results in unacceptable errors. To accomplish this, we are developing the PANX code (Purdue - Argonne - Northwestern - Xi'an) to implement a 2D/1D VNM transport method without spatial homogenization employing spherical harmonics and diffusion theory, respectively, in the radial and axial directions. In the preceding paper, we employed diffusion theory in conjunction with finite element methods to implement heterogeneous VNM nodes, in which the problem of fuel-coolant smearing at axial interfaces, found in existing 2D/1D approximations, is eliminated. In this work, we generalize the resulting VNM nodes to couple spherical harmonics in the x-y plane with axial diffusion theory. We then apply the method to the three-dimensional C5G7 benchmark problem. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

A Variational Nodal Approach to 2D/1D Pin Resolved Neutron Transport: I Diffusion Theory

The variational nodal method (VNM) as implemented in the VARIANT code has been widely employed to fast reactor cores, but much less so for thermal reactors. In what follows we report the initial phase of development of the PANX code (Purdue - Argonne - Northwestern - Xi'an) for three-dimensional pin resolved PWR transport calculations. In two dimensions, prototype modifications of VARIANT using finite elements to obtain heterogeneous nodes and quasi-reflected interface conditions to reduce the high angular-order interface conditions, have been successfully employed to the 2D C5G7 benchmark. In three dimensions, however, the polynomial interface conditions employed in VARIANT smear fuel and coolant at each axial interface, leading to lack of convergence as the axial mesh is refined. This problem is also encountered in the 2D/1D approximations in MOC based methods. The intent of this work is to eliminate these interface homogenization effects from VNM calculations. For clarity, in this work we examine these phenomena in diffusion theory. The next stage of development will incorporate spherical harmonic trial functions to achieve more accurate transport solutions. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Improved Variable-Transformation Algorithm for the Hexagonal Variational Nodal Method

Hexagonal Variational Nodal Method (VNM) has been widely employed in fast reactor core calculation codes. It provides the weak form solution of the second order transport equation. It employs a direct three-dimensional expansion instead of the transverse integration technique, with the singular term caused by transverse integration over the hexagonal nodal avoided. However, as the increasing Pn expansion order in angle, the number of moments increases dramatically, which significantly increases the requirements in both memory and computational time. As a consequence, Yang proposed the Variable-Transformation Algorithm. It firstly used the reverse Cuthill-McKee (RCMK) reordering method to transform the response matrices into block-diagonal. Then, for the unreducible matrices, orthogonal transformation was applied, after which the matrices were sparse enough to be reduced by using the RCMK reordering once more. In this summary, based on a newly developed hexagonal VNM code NECP-VioletHex, the Variable Transformation Algorithm was investigated and improved by adopting more constraints to the basis functions to further take advantage of the symmetric properties of the hexagonal-z nodes. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Development of a new casting method to fabricate U–Zr alloy containing minor actinides

Metal fuel slugs of U–Zr alloys for a sodium-cooled fast reactor (SFR) have conventionally been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents, such as Am, are problematic in a conventional injection casting method. As an alternative fabrication method, low pressure gravity casting has been developed. Casting soundness, microstructural characteristics, alloying composition, density, and fuel losses were evaluated for the following as-cast fuel slugs: U–10 wt% Zr, U–10 wt% Zr–5 wt% RE, and U–10 wt% Zr–5 wt% RE–5 wt% Mn. The U and Zr contents were uniform throughout the matrix, and impurities such as oxyen, carbon, and nitrogen satisfied the specification of total impurities less than 2,000 ppm. The appearance of the fuel slugs was generally sound, and the internal integrity was shown to be satisfactory based on gamma-ray radiography. In a volatile surrogate casting test, the U–Zr–RE–Mn fuel slug showed that nearly all of the manganese was retained when casting was done under an inert atmosphere.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

This report presents the performance characteristics of two “two-stage” fast spectrum fuel cycle options proposed to enhance uranium resource utilization and to reduce nuclear waste generation. One is a two-stage fast spectrum fuel cycle option of continuous recycle of plutonium (Pu) in a fast reactor (FR) and subsequent burning of minor actinides (MAs) in an accelerator-driven system (ADS). The first stage is a sodium-cooled FR fuel cycle starting with low-enriched uranium (LEU) fuel; at the equilibrium cycle, the FR is operated using the recovered Pu and natural uranium without supporting LEU. Pu and uranium (U) are co-extracted from the discharged fuel and recycled in the first stage, and the recovered MAs are sent to the second stage. The second stage is a sodium-cooled ADS in which MAs are burned in an inert matrix fuel form. The discharged fuel of ADS is reprocessed, and all the recovered heavy metals (HMs) are recycled into the ADS. The other is a two-stage FR/ADS fuel cycle option with MA targets loaded in the FR. The recovered MAs are not directly sent to ADS, but partially incinerated in the FR in order to reduce the amount of MAs to be sent to the ADS. This is a heterogeneous recycling option of transuranic (TRU) elements

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Value of Information in Fuel Cycle Decision-making

Transitions between nuclear fuel cycles with uncertain key parameters (e.g., SNF and HLW disposal costs, fast reactor capital costs) have been examined using a decision-making under uncertainty methodology. The methodology explicitly incorporates uncertainties in key parameters into the decision-making process by constructing a stochastic model and embedding uncertainties as bifurcations in the decision-tree. The framework of this methodology allows for interpretation using a real options valuation or game theoretic approach, reviewed in. The work presented here extends the decision-making under uncertainty framework to calculate the value of information, which may be used to inform a game against nature. These games against nature have been applied to resource allocation problems, where scarce resources are allocated to maximize expected payoffs. More broadly, the value of developing alternative technologies may be considered under the same framework, where an investment portfolio for technology deployment may be developed. Here, the value of information is taken as the value of reducing uncertainty in key parameters, and these values may be substituted for the traditional payoff matrix when considering allocating scarce resources. Games against nature may provide a unique perspective on R and D funding decisions when considering transitions between nuclear fuel cycles. By determining the value of reducing uncertainties (of gaining information on probability distributions) scarce resources may be allocated in such a way to maximize their utility. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Effects of Chemical Wastes on Concrete through Spectroscopic and Computational Models - 18405

Liquid mercury was used as a coolant for early fast reactors, as catalyst in the dissolution of uranium alloy fuels, lithium isotope separation and other processes. Cesium is produced during nuclear fission reactions within the reactor. The interaction between chemical contaminants and solid structures can cause structural damages (e.g. dislocations) and consequently modify the material's mechanical properties. Another problem associated with the interaction of chemical contaminants and metal structures is corrosion pitting. Mercury intrusion porosimetry (MIP) has been utilized to investigate the pore structure of the cement material. The overall goal of this project is to determine the physiochemical interactions of Hg and Cs to steel and concrete using spectroscopic analysis: - Evaluate the physicochemical interactions of Hg and Cs to steel and concrete using spectroscopic techniques under standard conditions; - Evaluate the impact of environmental conditions (temp., salt content, pH, and time) on the physicochemical interactions of Hg and Cs to steel and concrete using spectroscopic techniques. Conclusion and Future Work: Adsorption process depends on the surface morphology. Surface morphology analysis demonstrated that concrete samples had a heterogenous surface. Adsorption of Mercury and Cesium are more favorable onto heterogeneous regions such as edges, grooves, and wrinkles. The mechanism for cation removal by concrete includes surface complexation and ion exchange. Overall adsorption of Mercury was more favorable than Cesium onto concrete. Continuing investigations will be focusing on application considerations such as multicomponent interactions and desorption studies.

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

Design and testing of a combined neutron and gamma assay system for measuring fissile material in fuel and waste items - 15340

A combined neutron and gamma ray assay system for the measurement of fissile material in both fuel and waste items has been designed, modelled, developed and tested. The purpose of the system is to measure and characterize a wide range of fissile materials including uranium, plutonium, americium and mixed oxide fuel arising from a variety of processes and facilities across the United Kingdom. Dounreay Site Restoration Limited at Dounreay currently stores these items form the United Kingdom Nuclear Decommissioning Authority's un-irradiated fuel inventory. This inventory comprises historic materials from ex-United Kingdom Atomic Energy Authority nuclear fuel projects and experiments together with fast reactor fuel materials and some commercial fuel items as delivered to Dounreay for recovery. The combined assay system has been designed and constructed to support the operation of a fuel processing facility, which will be operated at Dounreay to treat these historic legacy fuel materials, making them suitable for long term storage. Appropriate characterization of the fissile material is required for safety, criticality control and nuclear materials safeguards purposes prior to final disposition. The system consists of both gamma ray and neutron measurement components. The results of a neutron measurement are combined with the gamma ray isotopic ratio measurement results to generate the assay result for the fissile material being characterized. The gamma ray component of the system employs a safeguards quality high purity Germanium detector with electro-mechanical cooling and the measurements of fissile material containers are made in a shielded chamber with a tin and copper graded lining. Containers of fissile material are rotated during the gamma ray measurement and a variable automatic steel collimator aperture mechanism is employed to adjust the measured count rate to control detector dead time. Gamma ray spectra are analysed using the plutonium and uranium isotopic ratio code PC/FRAM. The gamma ray system can also be configured to perform a direct assay of small quantities of uranium bearing material. For this assay measurement a series of cadmium filters can be positioned in front of the detector to reduce the contribution from 59 keV gamma rays from Am-241. In addition to gamma ray isotopic measurements, both active and passive neutron assay is performed by the system. The neutron-measuring component consists of a thick walled cylindrical polyethylene chamber, which provides neutron moderation for 30 one-inch diameter He-3 detector tubes located in re-entrant channels in the polyethylene moderator. Two removable polyethylene-inserts incorporating lower plug assemblies are provided, one with and one without a cadmium lining. Also two top plug units are provided, again one with cadmium and one without a cadmium liner. The Neutron Subsystem can be operated both as a passive and as an active neutron coincidence counter. In both fast and thermal active neutron mode, the neutron component employs two americium-beryllium (Am-Be) alpha-n neutron sources, one positioned in the top and the other in the bottom plug unit. The neutron sources are removed when the counter is operated in passive mode. Passive neutron detection efficiency and active neutron performance has been modelled using the MCNP Monte Carlo computer simulation code as part of the design process. In the absence of fissile samples, operation of the gamma ray component of the system has been tested initially using calibrated Eu-152 and Ba-133 gamma ray sources and the use of archived spectra from fissile material measurements. Initial testing of the neutron system has employed a calibrated Cf-252 spontaneous fission neutron source. Neutron testing has been confirmed using further MCNP simulations. The results of test measurements with the system are presented and compared with the results of MCNP measurement simulations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

A Design and Fabrication Options Trade Study for Metallic Fuel without Internal Sodium Bonding

Nuclear fuels using alloys of uranium, or metallic fuels, have many beneficial properties. The classical metallic fuel design uses a loose fitting cylindrical “slug” of fuel placed inside stainless-steel cladding tubes where the gap is filled with sodium. This sodium bond is liquid at operating temperature and conducts heat from the slug to the cladding, especially in early life before fuel swells into contact with the cladding. Despite the benefits of sodium bonding, there is a desire to develop metallic fuel technologies without it chiefly to reduce chemical reaction hazards in spent fuel storage from sodium fast reactors operating on once-through fuel cycles. Elimination of the sodium bond may also help unlock potential benefits for fuel fabrication, reactor neutronics, and compatibility with other types of reactors. Creating a sodium-free metallic fuel revolves around the problem of manufacturing fuel slug geometries which are in close contact with the cladding at beginning of life to facilitate heat transport while alleviating fuel-cladding chemical interactions (FCCI) at this interface and providing enough free volume to accommodate fuel swelling. Accelerating development and qualification of this fuel system will require careful selection of design and manufacturing options. To this end, a design trade-off study was performed to evaluate candidate options. Several design and manufacturing options were assessed, weighted, scored, and ranked with respect to fabrication, normal reactor operation, off-normal scenarios, and back-end considerations. This effort was performed both for “baseline” needs, which represented a once-through fuel cycle at temperatures and burnups known to be viable for sodium-bonded metallic fuel, and for “enhanced” needs to represent opportunities for closed fuel cycles and/or more aggressive temperatures/burnups. The outcomes of this study prioritized a baseline technology using U-Zr alloy with additives to mitigate FCCI, produced in annular slug geometry by continuous casting, clad in austenitic stainless-steel alloy, and followed by a final step to swage the cladding down to close the gap. This study prioritized an enhanced fuel technology using U-Mo alloy, also produced by continuous casting into an annular geometry, followed by coating/plating with an FCCI barrier on the slug, again with a final step to swage the cladding diameter down using oxide dispersion strengthened steel. It was noted that development of the enhanced fuel technology would entail more risk, thus U-Zr alloy was put forth as a backup to U-Mo if challenges are encountered with FCCI barriers, and advanced ferritic/martensitic steels are put forth as a backup to oxide dispersion strengthened steels if swaging and welding are found unworkable.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

An investigation of the phase behaviors for quaternary U-Nb/Mo-Ti-Zr metallic fuel alloys

We report quaternary fuel alloys containing U, Nb/Mo, Ti, and Zr are proposed as fuel candidates for sodium-cooled fast reactors (SFRs). In this work, two Nb-bearing alloys, i.e., U-NT5Z (U-2.5Nb-2.5Ti-5.0Zr in wt%) and U-NT7Z (U-1.5Nb-1.5Ti-7.0Zr in wt%), and two Mo-bearing alloys, i.e., U-MT5Z (U-2.5Mo-2.5Ti-5.0Zr in wt%) and U-MT7Z (U-1.5Mo-1.5Ti-7.0Zr in wt%) were characterized and compared. The characterization techniques were differential scanning calorimetry (DSC), X-ray powder diffraction (XRD), and scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/EDS). DSC was performed to obtain the transition behaviors, and XRD and SEM/EDS were applied for phase identification. The results were combined to obtain the solid-state phase transitions between 500 °C and 850 °C for the alloys. It is found that the Nb-bearing alloys comprise similar phase transition behaviors as the Mo-bearing alloys. The phase transitions in U-NT5Z are a →γ at 608 °C and U 2 Ti → γ at 627 °C, which are ~40 °C higher than that of U-MT5Z. The phase transition in U-NT7Z is a → γ at 645 °C, and is 23 °C higher than that of U-MT7Z.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Effects of GdCl{sub 3} on U Electrochemical Properties in LiCl-KCl-UCl{sub 3}-GdCl{sub 3} Salt

Studies on pyro-processing technology have been stimulated in several countries including the United State, Russia, France, Japan, South Korea, and India. The main purpose of this technology is to treat the irradiated nuclear fuel, reduce volume of the nuclear waste, recycle actinides, and close the fast reactor fuel cycle. Therefore, ongoing researches and developments are expected. In pyro-processing, the most important process is an electro-refiner (ER) where uranium and minor actinides are separated on electrode surface while rare earth fission products are accumulated in LiCl- KCl eutectic salt. Due to the fact that the recovery of the U and actinide (An) materials are conducted via an electrolysis, many researchers have been investigating fundamental properties of U in LiCl- KCl system by using various electrochemical techniques. The resulting datasets are useful for effective system operation and development of safeguards in ER system. However, these values have been measured in pure UCl{sub 3} system in LiCl- KCl salt, which may not be practical to the real ER system owing to the fact that a variety of other fission products including An and lanthanide (Ln) are coexisting in the ER system. Therefore, in this work, gadolinium (Gd) was selected as a surrogate material for An or Ln materials in ER system. This is because that Gd has similar physical characteristics to Ln and An elements such as ionic size and reduction potential. By conducting electrochemical methods including cyclic voltammetry (CV), open circuit potential (OCP), and electrochemical impedance spectroscopy (EIS), effects of Gd coexistence on the electrochemical, thermodynamic, and kinetic properties of U were observed. It is hoped that this study will help us further understanding fundamental effect of the Ln/An effect on LiCl- KCl-UCl{sub 3}, yielding closer to the practical ER unit operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗