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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Preliminary design of control rods in the single-fluid double-zone thorium molten salt reactor (SD-TMSR)

Recent studies on Molten Salt Reactors (MSRs) showed that the excess reactivity at the beginning of the operation is large for many fueling strategies and must be compensated by a reactivity control system. The current work introduces a reliable safety system based on control rods in addition to the online feed system reactivity control in the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). Three different initial fissile loadings are considered: 233 U, reactor-grade Pu, and transuranic (TRU) elements as a startup fuel. We applied six different absorbing materials to investigate the main operational and safety parameters in the SD-TMSR: natural B 4 C, enriched B 4 C with 90% 10 B, HfB 2 , HfH 1.62 , Eu 2 O 3 , and Gd 2 O 3 . The present work focuses on control rod design, integral and differential control rod worth, shutdown margin, and shadowing effects at steady-state. We employed the SERPENT-2 Monte-Carlo code to calculate the reactivity worth and analyze the performance of the reactivity control system. We showed that 233 U and reactor-grade Pu startup cores maintain adequate shutdown margin with all considered absorbers. Finally, this report proposes a design of control rod clusters that compensate the excess reactivity of the SD-TMSR loaded with different initial fissile material.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Nuclear fuels for transient test reactors

Transient test reactors with the ability to test fissile specimens under extreme conditions have been crucial tools in the development of nuclear technologies. Less than 10 unique facility designs have ever been constructed, most of which remain operational today and still use the original nuclear fuel constructed for them more than 40 years ago. Historic fuel systems for transient test reactors vary in significant ways which have marked influences on reactor capabilities. Eventually, new fuel will be needed to support the longevity of transient test reactor missions. This paper reviews precedent transient reactor fuel systems in the context of their unique requirements. A few key conclusions are illustrated by comparing and contrasting these transient test reactors. Fuel composites which are mostly graphite can enable transient reactors with very high neutron fluence capability (>2E16 n/cm 2 ) and are amenable to longer “shaped” transients but cannot achieve pulses <10 ms in duration. Reducing the graphite-to-uranium ratio can yield a very narrow pulse capability but delivers less fluence and requires cores with considerably more fissile material. Designs based on uranium dioxide (UO 2 ) make use of readily available materials to create compact cores with narrow pulse width capabilities but with moderate neutron fluence capabilities (~2E15 n/cm 2 ). Uranium zirconium hydride (U-ZrHx) is a well-established fuel system for pulsing reactors which has been intermittently manufactured throughout the decades. U-ZrHx offers similar capabilities to UO 2 designs in terms of nuclear kinetics, but with about half the fluence capability (~1E15 n/cm 2 ). An evolution of the UO 2 system, termed “ternary ceramic” fuel, shows that dispersing UO 2 in zirconium oxide and calcium oxide can increase fluence capability greatly (~7E15 n/cm 2 ), but is not presently a commonly available fuel form. A unique composite of UO 2 and beryllium oxide (UO 2 -BeO) can be used to create a core with similar kinetics and compact core geometry as U-ZrHx designs, but with significantly higher fluence capability (~6E15 n/cm 2 ). Like ternary ceramic fuel, newly fabricated UO 2 -BeO would require reestablishing its historic manufacturing process which would be further complicated by the health hazards associated with beryllium. In conclusion, like most engineering problems, there is no perfect solution, but this paper outlines the advantages and disadvantages of candidate fuel options to help guide detailed evaluations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Enhancing thermal conductivity of UO 2 with the addition of UB 2 via conventional sintering techniques

We report that uranium dioxide has been the primary fuel type used in light water reactors for more than 40 years and proven to be reliable and robust. However, the Fukushima-Daiichi nuclear accident has motivated new work evaluating fuels with characteristics promoting accident tolerance, including enhanced thermal conductivity. Recently, additives have been investigated to increase thermal conductivity, but research has been largely focused on non-fissile additions. This study investigated the use of fissile additives to not only increase the thermal conductivity but also increase the uranium loading. Uranium diboride was chosen as the additive for this study due to its promising corrosion behavior as well as its significantly higher thermal conductivity at 573 K (25 Wm -1 K -1 ) when compared to UO 2 (7 Wm -1 K -1 ). Uranium diboride powder was fabricated via the arc melting technique and a ball milling process prior to mixing with UO 2 in a 90/10 wt% UO 2 /UB 2 ratio. Green bodies were made using a uniaxial die and subjected to a traditional pressureless sintering technique at 2073 K in argon. Sintered samples were analyzed via laser flash analysis for thermal diffusivity and differential scanning calorimetry for specific heat capacity in order to calculate thermal conductivity. The samples displayed an increase of 36-55% in thermal conductivity between 323 K and 1273 K when compared to the benchmark samples (pure UO 2 ) as reported in open literature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

The effect of energy resolution on the extraction of information content from gamma-ray spectra

Rapid unambiguous identification of illicit radioactive materials is a matter of international concern that has yet to be fully attained. This is particularly true for shielded fissile materials in the presence of background radiation and routinely encountered benign radionuclides. Here, we present a systematic treatment of the effect of detector resolution on the identification of nuclides in simple and complex shielded γ-ray spectra. Case studies of some problematic spectra with 1000 counts suggest that, at moderate count rates, near-unambiguous identification (<1% probability of misidentification) of potential fissile sources requires a detector with an energy resolution of at least 1–2% at 662 keV. While somewhat limited in scope, the study establishes a quantitative basis for the comparative evaluation of detectors of comparable efficiency.

42 ENGINEERING↗
DOE OSTI2016

Low-resolution gamma-ray spectrometry for an information barrier based on a multi-criteria template-matching approach

Gamma-ray spectrometry has been successfully employed to identify unique items containing special nuclear materials. Template information barriers have been developed in the past to confirm items as warheads by comparing their gamma signature to the signature of true warheads. Their development has, however, not been fully transparent, and they may not be sensitive to some relevant evasion scenarios. Here, we develop a fully open template information barrier concept, based on low-resolution measurements, which, by design, reduces the extent of revealed sensitive information. The concept is based on three signatures of an item to be compared to a recorded template. The similarity of the spectrum is assessed by a modification of the Kolmogorov–Smirnov test to confirm the isotopic composition. The total gamma count rate must agree with the template as a measure of the projected surface of the object. In order to detect the diversion of fissile material from the interior of an item, a polyethylene mask is placed in front of the detector. Neutrons from spontaneous and induced fission events in the item produce 2.223 MeV gamma rays from neutron capture by hydrogen-1 in the mask. This peak is detected and its intensity scales with the item's fissile mass. This analysis is based on MCNP Monte Carlo simulations of various plutonium configurations suggests that this concept can distinguish a valid item from a variety of invalid ones. The concept intentionally avoids any assumptions about specific spectral features, such as looking for specific gamma peaks of specific isotopes, thereby facilitating a fully unclassified discussion. Finally, by making all aspects public and allowing interested participants to contribute to the development and benchmarking, we enable a more open and inclusive discourse on this matter.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Fast-neutron multiplicity counter for active measurements of uranium oxide certified material

We developed a fast-neutron multiplicity counter (FNMC) based on stilbene and EJ-309 organic 11 scintillators. The system can detect and discriminate correlated photon and neutron multiplets emitted by fission 12 reactions. We used the system to estimate the fissile mass of uranium oxide samples in active interrogation mode at the 13 Zero Power Physics Reactor of Idaho National Laboratory (INL). Two sets of certified reference material (CRM) 14 samples were characterized. The U-235 enrichment of the first set is constant at 93.2 wt%, and the UO 2 mass ranges 15 between 0.5 and 4 kg. The second set includes samples of increasing enrichment (from 20 wt% to 97 wt%) and constant 16 UO 2 mass of 230 g. We used two AmLi sources to induce fission reactions in the samples. Despite the intense gamma17 ray background of the UO 2 and interrogating sources, the system could measure induced fission neutrons emerging from 18 the interrogated samples without additional shielding surrounding the sample and only relying on pulse shape 19 discrimination to classify neutron and gamma-ray pulses. Here, the overall neutron count rate and time-correlated counts are 20 well correlated with the sample fissile mass. We also proved that CRM samples can be used to build a calibration curve 21 to assay the U-235 mass of unknown samples of different mass, geometry and enrichment, with an average bias error of 22 8%, for U-235 mass higher than 390 g.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Sub-critical assembly die-away analysis with organic scintillators

Determining the reactivity of systems containing fissile material is essential for verifying safety, and possession and enrichment levels. The goal of this experiment was to characterize a subcritical nuclear fuel assembly irradiated by a pulsed neutron generator through differential die-away analysis. Similar characterizations are commonly performed with thermal neutron detectors, but these detectors are bulky and cumbersome, the materials are not widely available, and timing can be clouded by the required thermalization time. In this work, an organic scintillator capable of detecting both neutrons and photons was used to measure the die-away profiles of both particles. These time profiles were then used to estimate the reactivity of the assembly. Here, the calculated reactivity estimates compare favorably to simulated estimates, and therefore provide a useful tool for fissile material characterization.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Spent fuel nondestructive assay integrated characterization from active neutron, passive neutron, and passive gamma

Spent nuclear fuel comprises a wide range of irradiated isotopic material compositions, and characterization through nondestructive measurements is beneficial in verifying declared parameters before the fuel is placed in storage, final disposal, and/or reprocessed. This paper discusses results from three nondestructive assay instruments, including passive gamma, passive neutron, and active neutron methods, that measured fifty spent fuel assemblies at the Clab interim storage facility in Sweden. Here, integrated analysis of the measurements from the three different instruments allowed parametric assessments of cooling time, burnup, neutron multiplication, fissile mass, initial enrichment, and decay heat of each individual fuel assembly. Passive gamma measurements were found to be the most beneficial in predicting cooling time, passive neutron for determining burnup, active neutron in estimating initial enrichment, and both passive and total neutron for multiplication correlations. Fissile mass was best estimated using any combination of any two of the instruments such that corrections for isotopic changes in the fuel could be accounted for with the first set of measurements and multiplication of the assembly in the second. In conclusion, the nondestructive assay technologies demonstrated through this effort enhance the characterization of spent nuclear fuel assemblies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Improved evaluation of safeguards parameters from spent fuel measurements with the Differential Die-Away (DDA) instrument

The Differential Die-Away (DDA) technique is a highly sensitive non-destructive assay method for characterizing and detecting the presence of fissile material within an item of interest. DDA utilizes a series of pulses from a neutron generator (NG) to actively interrogate an item of interest. The die-away time of the neutron population induced by this active interrogation and the integral of the total differential die-away signal can be used to characterize items such as nuclear waste drums and spent nuclear fuel assemblies. In this work, Los Alamos National Laboratory (LANL) conceptualized, designed, and fabricated a DDA instrument that was deployed for field test measurements at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab) in Oskarshamn, Sweden. The instrument performed multiple static measurements at fixed locations and dynamic axial scans of 15 pressurized water reactor (PWR) and 10 boiling water reactor (BWR) spent fuel assemblies, collecting both passive and active measurement data. The static assays of the assemblies measured the differential die-away signal, die-away time, and total passive neutron emission rate to create calibration curves for the evaluation of assembly multiplication, burnup, initial enrichment, effective fissile mass, and total elemental plutonium mass. Each calibration curve was optimized by minimizing the relative root mean square error (RRMSE) of assembly assay results compared to declared assembly parameters. The same quantities were also measured with the axial scans, and the resulting data were applied in two ways: (1) in the creation of calibration curves to improve evaluation of the same safeguards parameters as static assays, and (2) for comparison to simulation. In most cases, across both PWR and BWR assemblies, axial scan data improved the estimation of the above parameters, quantified by decreasing the calibration curve RRMSE. These axial scan results demonstrate the ability of the DDA instrument and analysis method to characterize spent PWR and BWR fuel as well as, or better than, a static assay of the same assembly. Furthermore, the DDA instrument’s unique ability to obtain both active and passive data in a single, axial scan of an entire spent fuel assembly represents a more efficient and accurate way of assaying spent fuel for verification purposes. These results represent a significant advancement for characterizing spent nuclear fuel compared to current technologies.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Realistic application of short-lived fission product delayed neutron, gamma-ray analysis for simultaneous nondestructive trace quantification of U, Pu mixtures on cellulose swipes

Detection and characterization of fissile traces are of interest to the international nuclear nonproliferation community, including the International Atomic Energy Agency. Pre-inspection check samples are analyzed by neutron activation analysis at the High Flux Isotope Reactor operated by the Oak Ridge National Laboratory under the umbrella of the IAEA Network of Analytical Laboratories. The simultaneous quantification of U and Pu mixtures was accomplished using the combined delayed neutron (DN) delayed gamma-ray (DG) method to analyze cellulose swipes with actinide loading <1ng in a blind field trial. The total fissile quantity was measured by the DN counts and the relative proportions of U, Pu were determined by calibration of the 104 Tc / 141 Ba fission product count ratio using known mixtures. The DNDG method demonstrated high accuracy in flagging the presence of 239 Pu in uranium down to <100 pg mass loading. In conclusion, peak significance tests helped to control false positive Pu flagging and simultaneous quantification of U and Pu loading was accomplished on samples that passed the significance tests.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Validation and demonstration of the AEFC as a practical safeguards tool for inventory verification

The Advanced Experimental Fuel Counter (AEFC) is a nondestructive assay (NDA) instrument designed to determine the residual fissile mass in irradiated fuel assemblies for safeguards verification purposes. This is done by actively interrogating an assembly with a neutron source and measuring the total (Singles) and correlated (Doubles) neutron count rates resulting from induced fissions in the irradiated nuclear fuel and relating those rates to the residual fissile mass using calibration curves. Comprehensive NDA measurements of the irradiated fuel inventory at Israeli Research Reactor 1 (IRR-1) were taken with the AEFC to validate a set of previously developed calibration curves. During the campaign, measurements were acquired of 32 standard fuel assemblies and three control assemblies in just nine days. This is a significant majority of the research reactor's irradiated fuel inventory and the largest data set gathered by the AEFC to date. Many of the fuel assemblies measured during the campaign had much shorter cooling times than those assemblies previously measured with the instrument. Calibration curves developed from previous AEFC deployments were used to determine the residual 235 U mass in the measured standard fuel assemblies. The results of the campaign demonstrated that the AEFC can be used to estimate the 235 U mass remaining in a large number of irradiated fuel assemblies with 1%–5% uncertainty in a reasonable amount of time despite operating in a high dose environment.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2011

Detection of anomalous reactor activity using antineutrino count evolution over the course of a reactor cycle

This paper analyzes the sensitivity of antineutrino count rate measurements to changes in the fissile content of civil power reactors. Such measurements may be useful in IAEA reactor safeguards applications. We introduce a hypothesis testing procedure to identify statistically significant differences between the antineutrino count rate evolution of a standard 'baseline' fuel cycle and that of an anomalous cycle, in which plutonium is removed and replaced with an equivalent fissile worth of uranium. The test would allow an inspector to detect anomalous reactor activity, or to positively confirm that the reactor is operating in a manner consistent with its declared fuel inventory and power level. We show that with a reasonable choice of detector parameters, the test can detect replacement of 82 kg of plutonium in 90 days with 95% probability, while controlling the false positive rate at 5%. We show that some improvement on this level of sensitivity may be obtained by various means, including use of the method in conjunction with existing reactor safeguards methods. We also identify a necessary and sufficient minimum daily antineutrino count rate and a maximum tolerable background rate to achieve the quoted sensitivity, and list examples of detectors in which such rates have been attained.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Detection of special nuclear material from delayed neutron emission induced by a dual-particle monoenergetic source

Detection of unique signatures of special nuclear materials is critical for their interdiction in a variety of nuclear security and nonproliferation scenarios. We report on the observation of delayed neutrons from fission of uranium induced in dual-particle active interrogation based on the {sup 11}B(d,n γ){sup 12}C nuclear reaction. Majority of the fissions are attributed to fast fission induced by the incident quasi-monoenergetic neutrons. A Li-doped glass–polymer composite scintillation neutron detector, which displays excellent neutron/γ discrimination at low energies, was used in the measurements, along with a recoil-based liquid scintillation detector. Time-dependent buildup and decay of delayed neutron emission from {sup 238}U were measured between the interrogating beam pulses and after the interrogating beam was turned off, respectively. Characteristic buildup and decay time profiles were compared to the common parametrization into six delayed neutron groups, finding a good agreement between the measurement and nuclear data. This method is promising for detecting fissile and fissionable materials in cargo scanning applications and can be readily integrated with transmission radiography using low-energy nuclear reaction sources.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

Technoeconomic Design Optimization for Fast Reactors. Part II: Impact of Technoeconomic Constraints on Optimal Design

There is a current drive toward optimizing reactors, particularly small/micro reactors to minimize cost and maximize performance. Previous work has investigated the development of technoeconomic workflows for the design optimization of pool-type fast reactors that aim to deploy into district energy grids. Initial scoping studies verified that the workflow was capable of capturing design trends throughout a variety of design configurations and problem formulations while remaining sufficiently flexible. In this paper, this methodology is applied to understand how cost functions and technoeconomic constraints can drive optimal reactor design. Specifically, the UPu10Zr-fueled fast reactor model from Part I is adapted to include changes in the fissile content limits, control rod worth limits, control rod drive cost, and assumed fuel form. In the case of constraint relaxation at fixed power (fissile content and control rod worth limits), cost sensitivities of 5% to 10% were uncovered. Multi-objective optimization at varying reactor power levels with individualized control rod drives for each assembly (as opposed to one operational and one safety drive) increased cost by $\$10$ to $\$25$ million and substantially altered the optimal core geometry, favoring geometries with substantially fewer control rod placements relative to baseline optimization. Finally, a multi-objective optimization was performed at varying power levels with the fuel form overhauled to metallic, high-assay low-enriched uranium–based U10Zr with more refined fuel cost models. In the case of uranium fueling, the costs increased by at least $50 million relative to the baseline case. Furthermore, economic fuel zoning and lower reactivity swing cores were recovered. Each case serves to demonstrate the value of applying technoeconomic workflows to initial reactor design scoping studies to better understand the trade-off for a proposed concept between different design options.

Argonne Reactor Computation (ARC) codes↗
DOE OSTI2021

Simulated Performance of the Micro-Pocket Fission Detector in the Advanced Test Reactor Critical Facility

The Micro-Pocket Fission Detector (MPFD) is a small-form–factor real-time fission chamber. MPFD performance has been simulated in the Advanced Test Reactor Critical Facility (ATRC), located at Idaho National Laboratory (INL). Here, the neutron and gamma-ray flux profiles and magnitudes were simulated using Monte Carlo N-Particle (MCNP) in the near-core B-8 irradiation position. These simulations were performed at 69 discrete axial locations inside the B-8 position for three separate orientations of the nearby hafnium outer shim control cylinders and at a power level of 700 Wth. The resulting neutron and gamma-ray flux values were used to determine the MPFD response for various fissile masses and detector gas pressures. The optimal gas-operating pressure was determined to be between 30 and 60 psig. The required fissile-layer mass was determined to be between 1–2 µg of 235 U. Additionally, the gamma-ray to fission-fragment interaction rate was determined to be 4.42 × 10 -3 with average energy deposition for gamma rays and fission fragments in 30 psig argon gas to be 1 keV and 9.5 MeV, respectively.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Supporting Technology for Chain of Custody of Nuclear Weapons and Materials throughout the Dismantlement and Disposition Processes

The ratification and ongoing implementation of the New START Treaty have been widely regarded as noteworthy global security achievements for both the Obama Administration and the Putin (formerly Medvedev) regime. But deeper cuts that move beyond the United States and Russia to engage the P-5 and other nuclear weapons possessor states are envisioned under future arms control regimes, and are indeed required for the P-5 in accordance with their Article VI disarmament obligations in the Nuclear Non-Proliferation Treaty. Future verification needs will include monitoring the cessation of production of new fissile material for weapons, monitoring storage of warhead components and fissile materials and verifying dismantlement of warheads, pits, secondary stages, and other materials. A fundamental challenge to implementing a nuclear disarmament regime is the ability to thwart unauthorized material diversion throughout the dismantlement and disposition process through strong chain of custody implementation. Verifying the declared presence, or absence, of nuclear materials and weapons components throughout the dismantlement and disposition lifecycle is a critical aspect of the disarmament process. From both the diplomatic and technical perspectives, verification under these future arms control regimes will require new solutions. Since any acceptable verification technology must protect sensitive design information and attributes to prevent the release of classified or other proliferation-sensitive information, non-nuclear non-sensitive modalities may provide significant new verification tools which do not require the use of additional information barriers. Alternative verification technologies based upon electromagnetic and acoustics could potentially play an important role in fulfilling the challenging requirements of future verification regimes. For example, researchers at the Pacific Northwest National Laboratory (PNNL) have demonstrated that low frequency electromagnetic signatures of sealed metallic containers can be used to rapidly confirm the presence of specific components on a yes/no basis without revealing classified information. PNNL researchers have also used ultrasonic measurements to obtain images of material microstructures which may be used as templates or unique identifiers of treaty-limited items. Such alternative technologies are suitable for application in various stages of weapons dismantlement and often include the advantage of an inherent information barrier due to the inability to extract classified weapon design information from the collected data. As a result, these types of technologies complement radiation-based verification methods for arms control. This article presents an overview of several alternative verification technologies that are suitable for supporting a future, broader and more intrusive arms control regime that spans the nuclear weapons disarmament lifecycle. The general capabilities and limitations of each verification modality are discussed and example technologies are presented. Potential applications are defined in the context of the nuclear material and weapons lifecycle. Example applications range from authentication (e.g., tracking and signatures within the chain of custody from downloading through weapons storage, unclassified templates and unique identification) to verification of absence and final material disposition.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2016

Market-based Policies for Non-Proliferation

Access to weapons-useable fissile material—highly enriched uranium and separated plutonium—is the primary technical barrier to nuclear weapons development. Nuclear non-proliferation policy is largely concerned with the secure distribution and use of these materials and the technologies that produce them. To the extent that both materials and technologies are traded in markets, the problem may be viewed as one of regulating markets to give an acceptable set of allocative outcomes. While a laissez-faire approach to these markets was rejected long ago, most current policy is based on prescribed or proscribed behavior rather than market incentives. “Market-based” non-proliferation policies would act through incentives in the nuclear materials and technology markets rather than by prescribing outcomes or behavior. Possible polices of this sort include subsidized enrichment services, credits to impute value to spent nuclear fuel, and new markets for special scarce nuclear materials. While market-based policies will not replace command-and-control policies entirely, they could lead to efficient regulation which suffices for most market participants, reserving special, more discriminatory instruments for the few hard cases.

nulcear fuel markets, nonproliferation, SWU↗