Fuel Cycle Options Research Articles

Preliminary conceptual study of engineering-scale pyroprocess demonstration facility

The development of an effective management technology of spent fuel is important to enhance environmental friendliness, cost viability and proliferation resistance. In Korea, pyroprocess technology has been considered as a fuel cycle option to solve the spent fuel accumulation problems. PRIDE (PyRoprocess Integrated inactive DEmonstration facility) has been developed from 2007 to 2012 in Korea as a cold test facility to support integrated pyroprocessing and an equipment demonstration, which is essential to verify the pyroprocess technology. As the next stage of PRIDE, the design requirements of an engineering-scale demonstration facility are being developed, and the preliminary conceptual design of the facility is being performed for the future. In this paper, the main design requirements for the engineering-scale pyroprocess demonstration facility were studied in the throughput of 10tHM a year. For the preliminary conceptual design of the facility, the design basis of the pyroprocess hot cell was suggested, and the main equipment, main process area, operation area, maintenance area, and so on were arranged in consideration of the effective operation of the hot cells. Also, the argon system was designed to provide and maintain a proper inert environment for the pyroprocess. The preliminary conceptual design data will be used to review the validity of the engineering-scale pyroprocess demonstration facility that enhances both safety and nonproliferation.

Accelerator Driven Systems (ADSs) for nuclear transmutation

The paper summarizes the rationale and the objectives of Partitioning and Transmutation (P&T) research and technology development activities. It then provides overviews of both the various nuclear fuel cycle options and the transmutation technologies (with particular emphasis on ADS), as well as a brief worldwide summary of the main research and technology development efforts in the field of ADS.

When Is the Simple Radiotoxicity Approach Useful for Advanced Fuel Cycle System Assessments Given the Existence of Complex Performance Dose Assessments?

Engineers often face the general question of which approximations are appropriate for a given analytical task. In particular, when is a simpler model useful if a more complex model also exists? This paper explores this question in the domain of radiotoxicity relative to geologic disposal performance dose assessments.Criterion 1 requires that the simpler approach, radiotoxicity, must be calculated correctly. The concept of ingestion radiotoxic inventories is analogous to the inventory of toxic chemicals in other industries. From a decision analysis perspective, it is also somewhat analogous to the nuclear reactor safety concept of “passive safety.” This paper explains some of the issues in calculating radiotoxicity, motivated by the author’s observations of errors in the literature: not accounting for radioactive progeny, misunderstanding natural “ore,” and focusing on transuranic (TRU) isotopes without adequate attention to actinide decay products. For example, Th/233U fuel cycles do have lower amounts of TRU isotopes than U/239Pu fuel cycles, but that does not necessarily mean lower long-term hazard.Criterion 2 requires that the uncertainties in the more complex approach, performance dose assessments, must raise issues for the assessments’ intended purposes - in which case, radiotoxic inventory may be of assistance until those issues are resolved. Performance dose assessments were developed for, and are legally the way to show, compliance with regulations, but the uncertainties are large. Less obvious is the degree to which dose assessments are applicable to other purposes - comparing fuel cycle options prior to site selection and showing the safety of a fuel cycle and waste management approach to the public. In the last sense especially, performance dose assessments are analogous to probabilistic risk assessments for nuclear reactor safety. The United States lacks a selected consensus site, selected fuel cycle approach (direct disposal versus recycling), and selected waste form. Thus, the paper does not intend to discuss all the issues with performance dose assessments but rather intends to focus on only those performance dose uncertainties that raise issues when comparing fuel cycles. Uncertainties associated with whether a generic geological environment is attractive or a specific location meets requirements are beyond the scope of this paper.Ingestion radiotoxicity correlates with heat, gamma, and inhalation radiotoxicity. Thus, options that are relatively high in ingestion radiotoxicity tend to be high in other parameters. Therefore, reduction in ingestion radiotoxicity means both that the potential source term for release is lower but also that one driving force for release (heat) is also lower. However, the most important time frames differ as heat is mainly an issue in decades and centuries after reactor discharge, but ingestion radiotoxicity is mainly an issue during longer time periods. Ingestion radiotoxicity points to the importance of actinides in long-term waste management, followed by specific fission products such as 99Tc, 129I, 93Zr, 135Cs, and 79Se. TRU isotopes were important in four of five generic geologic environments recently studied independently with used fuel disposal experts as were the same fission products as noted from radiotoxicity - in part because waste assessments must include multiple exposure pathways including human intrusion and drilling into the waste. Dose calculations that were made for used fuel disposal may be misleading if extrapolated to disposal of only the remnants of used fuel separation because the chemistry of waste will differ. Radiotoxicity can be misleading if used to strictly rank order among individual isotopes if a specific disposal option is well known. Hybrid approaches that incorporate radiotoxicity and features of full performance assessments may have value.

The application of systems engineering principles to the prioritization of sustainable nuclear fuel cycle options

We investigate the implementation of the principles of systems engineering in the U.S. Department of Energy’s Fuel Cycle Technologies (FCT) Program to provide a framework for achieving its long-term mission of demonstrating and deploying sustainable nuclear fuel cycle options. A fuel cycle “screening” methodology is introduced that provides a systematic, objective, and traceable method for evaluating and categorizing nuclear fuel cycles according to their performance in meeting sustainability objectives. The goal of the systems engineering approach is to transparently define and justify the research and development (R&D) necessary to deploy sustainable fuel cycle technologies for a given set of national policy objectives. The approach provides a path for more efficient use of limited R&D resources and facilitates dialog among a variety of stakeholder groups interested in U.S. energy policy. Furthermore, the use of systems engineering principles will allow the FCT Program to more rapidly adapt to future policy changes, including any decisions based on recommendations of the Blue Ribbon Commission on America’s Nuclear Future. Specifically, if the relative importance of policy objectives changes, the FCT Program will have a structured process to rapidly determine how this impacts potential fuel cycle performance and the prioritization of needed R&D for associated technologies.

Characterizing Limited-Separations Fuel Cycles Using Breed-and-Burn Reactors

A fuel cycle option is evaluated in which fuel bred in breed-and-burn (B&B) reactors is used to start up additional B&B reactors, with the fuel being recycled using limited-separations processes instead of full actinide reprocessing. This fuel cycle aims to minimize processing requirements and proliferation risk while still being able to achieve exponential growth and high uranium utilization. The neutron excess concept is applied to compute the starting fuel requirements of new B&B reactors, allowing fleet doubling times to be estimated. A simple analytic expression for doubling time is derived, which is applied to example B&B reactors using a hypothetical core composition. It is found that larger reactors are able to achieve shorter doubling times because of their smaller starter fuel requirements per unit power. Several variant fuel cycle configurations are examined, and their doubling times are computed.

Sustainability Features of Nuclear Fuel Cycle Options

The nuclear fuel cycle is the series of stages that nuclear fuel materials go through in a cradle to grave framework. The Once Through Cycle (OTC) is the current fuel cycle implemented in the United States; in which an appropriate form of the fuel is irradiated through a nuclear reactor only once before it is disposed of as waste. The discharged fuel contains materials that can be suitable for use as fuel. Thus, different types of fuel recycling technologies may be introduced in order to more fully utilize the energy potential of the fuel, or reduce the environmental impacts and proliferation concerns about the discarded fuel materials. Nuclear fuel cycle systems analysis is applied in this paper to attain a better understanding of the strengths and weaknesses of fuel cycle alternatives. Through the use of the nuclear fuel cycle analysis code CAFCA (Code for Advanced Fuel Cycle Analysis), the impact of a number of recycling technologies and the associated fuel cycle options is explored in the context of the U.S. energy scenario over 100 years. Particular focus is given to the quantification of Uranium utilization, the amount of Transuranic Material (TRU) generated and the economics of the different options compared to the base-line case, the OTC option. It is concluded that LWRs and the OTC are likely to dominate the nuclear energy supply system for the period considered due to limitations on availability of TRU to initiate recycling technologies. While the introduction of U-235 initiated fast reactors can accelerate their penetration of the nuclear energy system, their higher capital cost may lead to continued preference for the LWR-OTC cycle.

Economic Analysis of Different Nuclear Fuel Cycle Options

An economic analysis has been performed to compare four nuclear fuel cycle options: a once-through cycle (OT), DUPIC recycling, thermal recycling using MOX fuel in a pressurized water reactor (PWR-MOX), and sodium fast reactor recycling employing pyroprocessing (Pyro-SFR). This comparison was made to suggest an economic competitive fuel cycle for the Republic of Korea. The fuel cycle cost (FCC) has been calculated based on the equilibrium material flows integrated with the unit cost of the fuel cycle components. The levelized fuel cycle costs (LFCC) have been derived in terms of mills/kWh for a fair comparison among the FCCs, and the results are as follows: OT 7.35 mills/kWh, DUPIC 9.06 mills/kWh, PUREX-MOX 8.94 mills/kWh, and Pyro-SFR 7.70 mills/kWh. Due to unavoidable uncertainties, a cost range has been applied to each unit cost, and an uncertainty study has been performed accordingly. A sensitivity analysis has also been carried out to obtain the break-even uranium price (215$/kgU) for the Pyro-SFR against the OT, which demonstrates that the deployment of the Pyro-SFR may be economical in the foreseeable future. The influence of pyrotechniques on the LFCC has also been studied to determine at which level the potential advantages of Pyro-SFR can be realized.

Influence of the Monovalent M Cation on the M2+xTh2-xNdx(C2O4)5 nH2O Solid Solution Domain

An optimized fuel cycle option for Generation III/IV systems could be co-management of the minor actinides in an integrated closed fuel cycle. This approach implies separating these actinides from the fission products, and then converting them to solid forms to re-fabricate fresh fuel. Oxalate compounds are well known for this purpose. Co-management process is based on the oxalic co-precipitation of An(IV) and An(III) in a mixed oxalate as, for example, the hexagonal phase, M+2+xAnIV2-xAnIIIx(C2O4)5nH2O with M+ representing a singly-charged cation. A study of the Th-Nd-M+ hexagonal oxalate systems has been carried out in order to understand the charge compensation mechanisms implicated in the formation of mixed oxalates systems.

Chemical Pre-Treatment of used Fuel for Long–Term Storagea

Repository programs throughout the world have been slowed by the need for increased local public involvement in the siting and licensing process. The result has been an increase in the dry storage of used fuel at reactor sites and the potential that such storage may be extended for many decades, even centuries. While there are sound technical reasons to believe that dry storage can be conducted safely, there are increasing concerns that the ultimate transfer to either a future repository or a centralized separations plant may result in fractured cladding and serious handling issues, including criticality concerns. These concerns would be increased for higher burn-up fuels. Currently, various chemical pre-treatment processes under R&D for application to commercial used oxide fuel have been investigated at the laboratory scale as methods to simplify and increase the safety of the remaining stages of conventional solvent extraction processing. This includes advanced decladding methods and various oxidation/reduction processes designed to release volatile and semi-volatile fission products, produce finely divided uranium oxide powder, and ameliorate the subsequent nitric acid dissolution step. The paper examines the potential for combining several chemical and physical pre-treatment steps to minimize long-term concerns about safe transport of used fuel, possibly providing another option for future nuclear waste management. Laboratory data from both cold and hot testing will provide the basis for the evaluation. An example of a potential pre-treatment process includes shearing, advanced voloxidation and off-gas treatment, the possible mixing of the resulting uranium oxide with a secondary oxide, and densification and recanning in nitric acid-soluble storage containers for extended time periods. Chemical decladding may be feasible to replace shearing. Zirconium recycle may also be feasible, significantly reducing high level waste quantities. Both analytic and experimental data will be applied to the examination of this potential fuel cycle option.

Comparative study of different nuclear fuel cycle options: Quantitative analysis on material flow

As a nation develops its nuclear strategies, it must consider various aspects of nuclear energy such as sustainability, environmental-friendliness, proliferation-resistance, economics, technologies, and so on. A nuclear fuel cycle study could give convincing answers to many questions in regard to technical aspects. However, one nuclear fuel cycle option cannot be superior in all aspects. Therefore a nation must identify its top priority and accordingly evaluate all the possible nuclear fuel cycle options. For such a purpose, this paper examined four different fuel cycle options that are likely to be plausible under situation of Republic of Korea: once-through cycle, DUPIC recycling, thermal recycling using MOX fuel in PWR (pressurized water reactor), and SFR (sodium cooled fast reactor) employing fuel recycling by a pyroprocess. The options have been quantitatively compared in terms of resource utilization and waste generation based on 1 TWh electricity production at a “steady-state” condition as a basic analysis. This investigation covered from the front-end of the fuel cycles to the final disposal and showed that the Pyro-SFR recycling appears to be the most competitive from these material quantitative aspects due to the reduction of the required uranium resources and the least amount of waste generation.

Dynamic Simulations of Advanced Fuel Cycles

Nothing in life is static, so why compare fuel cycle options using only static, equilibrium analyses? Competitive industry looks at how new technology options might displace existing technologies and change how existing systems work. So too, our years of performing dynamic simulations of advanced nuclear fuel cycle options provide insights into how they might work and how one might transition from the current once-through fuel cycle. This paper summarizes those insights within the context of the 2005 objectives and goals of what was then the U.S. Advanced Fuel Cycle Initiative (AFCI). The intent here is not to compare options, assess options versus those objectives and goals, nor recommend changes to those objectives and goals. (The specific options change over time; the objective in this paper is to look for more generic insights.) We organize what we have learned from dynamic simulations in the context of the AFCI objectives for waste management, proliferation resistance, uranium utilization, and economics. Thus, we do not merely describe “lessons learned” from dynamic simulations but attempt to answer the “so what” question by using this context; i.e., how do the lessons learned matter relative to goals and objectives not just to technological observations? The analyses have been performed using the Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics (VISION). We observe that the 2005 objectives and goals do not address many of the inherently dynamic discriminators among advanced fuel cycle options and transitions thereof.

Recognising interdependencies in design of nuclear fuel cycle and transportation of SNF and HLW

Spent nuclear fuel (SNF) discharged from a nuclear power plant and high level radioactive waste (HLW) generated during reprocessing SNF are typically stored at the sites where they are generated, often for prolonged periods of time. Eventually, however, these materials must be transported off-site to an interim storage facility, a reprocessing plant, or directly to a deep geologic repository. This paper considers the interdependencies between nuclear fuel cycle options and the transportation system, and argues that both must be addressed as part of an integrated system. Two examples are presented to illustrate why this is important. The first draws from experience gained during development of the US Department of Energy’s (DOE) program for disposing of SNF and HLW in a proposed repository. In particular, decisions DOE made with regard to waste package design had profound implications on the viability of the transportation system. The second example relates to operational changes now underway at nuclear po...

Benefits and concerns of a closed nuclear fuel cycle

Nuclear power can play an important role in our energy future, helping to meet increasing electricity demand while at the same time decreasing carbon dioxide emissions. However, the nuclear fuel cycle in the United States today is unsustainable. The 1982 Nuclear Waste Policy Act establishes the U.S. Department of Energy as responsible for disposing of spent nuclear fuel (SNF) generated by commercial nuclear power plants operating in a “once-through” fuel cycle in a deep geologic repository located at Yucca Mountain, NV. However, unyielding political opposition to the Yucca Mountain site has hindered the commissioning process to the extent that the current administration has recently declared the site unsuitable. In light of this, the DOE is exploring other options, including closing the fuel cycle through reprocessing and recycling of spent nuclear fuel. The possibility of closing the fuel cycle is receiving special attention because of its ability to minimize the final high level waste package by separating and isolating the most long-lived components, as well as recovering additional energy value from the original fuel. Reprocessing and recycling of SNF can decrease the volume of waste stored by a factor of 4 and reduce the timeframe of storage from hundreds of thousands of years to thousands of years. Reprocessing and recycling technologies are, however, still very controversial because of the increased cost and proliferation risk reprocessing can present. Estimates of increases in the levelized cost of electricity with reprocessing range from about 10% to 50% due to large uncertainties in the financing, construction, and licensing of a new plant. Ultimately, the U.S. will need to compare each of these fuel cycle options with respect to sustainability, proliferation risk, commercial viability, waste management, and energy security to define the future of nuclear power.

Impact of PWR spent fuel variations on back-end features of advanced fuel cycles with tru-fueled VHTR

Alternative strategies are being considered as management option for current spent nuclear fuel transuranics (TRU) inventory. Creation of transmutation fuels containing TRU for use in thermal and fast reactors is one of the viable strategies. Utilization of these advanced fuels will result in transmutation and incineration of the TRU. The objective of this study is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled very high temperature reactor (VHTR) systems. The current effort is focused on prismatic core configuration operated under a single batch once-through fuel cycle option. IAEA’s nuclear fuel cycle simulation system (VISTA) was used to determine potential PWR spent fuel compositions. Additional composition was determined from the analysis of United States legacy spent fuel that is given in the Yucca Mountain Safety Assessment Report. A detailed whole-core 3-D model of the prismatic VHTR was developed using SCALE5.1 code system. The fuel assembly block model was based on Japan’s HTTR fuel block configuration. To establish a reference reactor system, calculations for LEU-fueled VHTR were performed and the results were used as the basis for comparative studies of the TRU-fueled systems. The LEU fuel is uranium oxide at 15% 235U enrichment. The results showed that the single-batch core lifetimes ranged between 5 and 7 years for all TRU fuels (3 years in LEU), providing prolonged operation on a single batch fuel loading. Transmutation efficiencies ranged between 19% and 27% for TRU-based fuels (13% in LEU). Total TRU material contents for disposal ranged between 730 and 808 kg per metric ton of initial heavy metal loading, reducing TRU inventory mass by as much as 27%. Decay heat and source terms of the discharged fuel were also calculated as part of the spent fuel disposal consideration. The results indicated strong potential of TRU-based fuel in VHTR.

A comparative study on recycling spent fuels in gas-cooled fast reactors

This study evaluates advanced Gas-cooled Fast Reactor (GFR) fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. A 600 MWth GFR was used for the fuel cycle analysis, and the equilibrium core was searched with different fuel-to-matrix volume ratios such as 70/30 and 60/40. Two fuel cycle scenarios, i.e., a one-tier case combining a Light Water Reactor (LWR) and a GFR, and a two-tier case using an LWR, a Very High Temperature Reactor (VHTR), and a GFR, were evaluated for mass flow and fuel cycle cost, and the results were compared to those of LWR once-through fuel cycle. The mass flow calculations showed that the natural uranium consumption can be reduced by more than 57% and 27% for the one-tier and two-tier cycles, respectively, when compared to the once-through fuel cycle. The transuranics (TRU) which pose a long-term problem in a high-level waste repository, can be significantly reduced in the multiple recycle operation of these options, resulting in more than 110 and 220 times reduction of TRU inventory to be geologically disposed for the one-tier and two-tier fuel cycles, respectively. The fuel cycle costs were estimated to be 9.4 and 8.6 USD/MWh for the one-tier fuel cycle when the GFR fuel-to-matrix volume ratio was 70/30 and 60/40, respectively. However the fuel cycle cost is reduced to 7.3 and 7.1 USD/MWh for the two-tier fuel cycle, which is even smaller than that of the once-through fuel cycle. In conclusion the GFR can provide alternative fuel cycle options to the once-through and other fast reactor fuel cycle options, by increasing the natural uranium utilization and reducing the fuel cycle cost.

“Used fuel” vectors and waste minimization strategies for VHTRS operating without refueling

Generation IV Very High Temperature Reactors (VHTRs) are well-known for their flexibility with respect to feasible fuel cycle options. In this paper, the LEU- and TRU-fueled VHTR configurations are analyzed accounting for their capabilities to attain an extended single-batch OTTO (Once-Through-Then-Out) mode of operation without intermediate refueling. The requirement of waste minimization is imposed as one of the design constraints defining possible system configurations and deployment strategies. The resulting “used fuel” vectors are examined considering anticipated disposal options as well as viability of fuel reprocessing. A Monte Carlo-deterministic analysis methodology has been implemented for coupled design studies of VHTRs with TRUs using the ORNL SCALE 5.1 code system. The developed modeling approach provides an exact-geometry 3D representation of the VHTR core details properly capturing VHTR physics. The presented analysis is focused on prismatic block core concepts for VHTRs. It is being performed within the scope of the U.S. DOE NERI project on utilization of higher actinides (TRUs and partitioned MAs) as a fuel component for extended-life VHTR configurations.

Nuclear power and the fuel cycle

No AccessNuclear power and the fuel cyclePublished Online:May 25, 2010pp 111-116https://doi.org/10.1504/IJNGEE.2010.033306PDF ToolsAdd to FavouritesDownload CitationsTrack Citations Share this article on social mediaShareShare onFacebookTwitterLinkedInReddit AboutAbstractDue to rising energy costs and climate concerns, nuclear power is once again being seriously considered as an energy source by several countries. This revival of nuclear power is closely linked with the choice of fuel cycles available, and the intentions of countries pursuing nuclear power are likely to be, correctly or incorrectly, judged by the choice of fuel cycle they make. The needs and constraints of the emerging nuclear powers may, however, be different from the expectations of a segment of the world community. If this potential growth in nuclear power is not to be stifled, it is imperative that a climate of mutual trust is developed respecting every country's right to develop peaceful uses of nuclear power without leading to an atmosphere of mistrust regarding the 'intentions' behind the pursuit of peaceful nuclear power. While it will be a near impossibility to completely decouple the peaceful uses of nuclear power from its more destructive applications, it is important that aspiring countries develop a clear and transparent process. Technology-supplier countries also need to develop and follow clear and consistent treaties and national policies, avoiding ad hoc country-specific arrangements. We review here the state of interest in nuclear power and current policies and discuss fuel cycle options that may pave the way for the future growth of nuclear power.Keywordsfuel cycle, policy, emerging nuclear powers, nuclear power, nuclear energy Next article FiguresReferencesRelatedDetails Volume 3Issue 22010 ISSN: 1742-4186eISSN: 1742-4194 HistoryPublished onlineMay 25, 2010 Keywordsfuel cyclepolicyemerging nuclear powersnuclear powernuclear energyPDF download

Prospective scenarios of nuclear energy evolution on the XXIst century over the world scale

Different world scenarios of nuclear energy development over the XXIst century are analyzed in this paper, by means of the EDF fuel cycle simulation code for nuclear scenario studies, TIRELIRE – STRATEGIE. Two nuclear demand scenarios are considered, and the performance of different nuclear strategies in satisfying these scenarios is analyzed and discussed, focusing on the maximum deployable capacity and the natural uranium consumption. Both thermal-spectrum systems (Pressurized Water Reactor, PWR, and High Temperature Gas-cooled Reactor, HTGR) and different designs of Fast Breeder Reactor (FBR) are investigated. A sensitivity analysis on the FBR deployment date, Breeding Gain and fuel cycle options is also presented.

Radionuclide inventories in the discharged fuels of PHWR-220, BWR-160, VVER-1000 and the conceptual ATBR-600 reactors – A case study

Radionuclides content in the discharged fuel of the conceptual thorium breeder reactor ATBR-600 has been assessed and compared against other thermal power reactors considered in Indian nuclear power programme. The contribution of actinides and the fission products inventories in the discharged fuels are separately estimated and assessed. The ATBR-600 reactor is suggested for closed fuel cycle option. The relatively large presence of the unspent plutonium would in fact be recycled. Nonetheless, the data has been presented in the event of operating ATBR-600 like other present day power reactors in a once through fuel cycle mode.

Proliferation Resistance of Americium Originating from Spent Irradiated Reactor Fuel of Pressurized Water Reactors, Fast Reactors, and Accelerator-Driven Systems with Different Fuel Cycle Options

The three most important americium isotopes, 241Am, 242mAm, and 243Am originate in the nuclear fuel of pressurized water reactors (PWRs), fast reactors (FRs), or accelerator-driven systems (ADSs) in a ratio of 241Am/243Am between ~0.45/0.55 to ~0.85/0.15. The content of 242mAm in the spent fuel of PWRs, FRs, and ADSs is relatively small and varies between 0.08 and 4.5%. Only by dedicated breeding in 241Am fuel and blanket assemblies could this 242mAm content be increased to ~7%. Only the isotope 241Am has a relatively high alpha-particle heat production whereas the isotopes 242mAm and 243Am have a relatively small alpha-particle heat production. All three americium isotopes are spontaneous fission neutron emitters.In this paper the different isotopic compositions of the three americium isotopes, 241Am, 242mAm, and 243Am are assembled for a number of fuel cycle strategies for PWRs, FRs and ADSs. Then, the critical masses, spontaneous fission neutron sources, and alpha-particle heat power of these different americium compositions are calculated. In a preignition analysis for gun systems and implosion systems, it is shown that only the implosion system would be applicable to the considered americium isotopic compositions. A subsequent thermal analysis with assumptions for the geometry and choice of materials of so-called hypothetical nuclear explosive devices (HNEDs) shows that the high alpha-particle heat power in the fissile reactor americium part would lead to such high temperatures that the surrounding chemical high explosives would melt and self-explode, and the americium metal would melt.Such HNEDs on the basis of reactor americium as fissile material would be technically unfeasible.