Zeev Shayer - Academia.edu (original) (raw)

Papers by Zeev Shayer

Journal of materials engineering and performance, Jun 14, 2024

Nuclear Science and Engineering, 1987

The sensitivity of the accuracy of the forward space and adjoint space calculation of a detector ... more The sensitivity of the accuracy of the forward space and adjoint space calculation of a detector response in a source driven system to the order of the angular quadrature used is investigated. It is found that for problems characterized by a localized source and extended detector distributions, the adjoint space calculation may require a significantly lower order of angular quadrature than the forward space calculation. The reciprocal situation prevails for problems characterized by an extended source and localized detector distributions.

The Pulsed Fast Neutron Analysis (PFNA) method utilizes the capability of high-energy neutrons to... more The Pulsed Fast Neutron Analysis (PFNA) method utilizes the capability of high-energy neutrons to penetrate deeply and stimulate the emission of characteristic gamma rays that can be detected and used to identify and image the emitting chemical elements. The PFNA method interrogates the object using a directional beam consisting of short bursts of fast neutrons from a pulsed deuteron accelerator, with the neutrons generated in a deuteron gas target. In order to study the performance of the PFNA system, a National Electrostatics Corporation (NEC) Pelletron accelerator was acquired and installed at the Science Applications International Corporation (SAIC) facility in Santa Clara, California. The NEC Pelletron accelerator produces 6- or 12-MHz pulses of up to 6-MeV deuterons with a beam current that will reach up to 100 [mu]A. There are several identifiable radiation sources during operation of the PFNA system: (1) neutron production target and deuteron beam stop; (2) neutrons and gamm...

Transactions of the American Nuclear Society, Dec 31, 1994

Most of the effort and research on the transmutation of radioactive waste have been focused on th... more Most of the effort and research on the transmutation of radioactive waste have been focused on the utilization of high neutron flux in advanced power reactors or n hybrid accelerator/subcritical systems. Some of the studies indicated that it might be not so economical to transmute {sup 90}Sr and {sup 137}Cs in these systems. Also these isotopes have a slight effect on the issues of geologic storage time because of their relatively short natural half-life ({approximately} 30 yr), but they might significantly affect the working protocol for a waste management period nearly comparable to the human life span. A novel approach to fission product transmutation is presented in this paper that is based on photon-neutron interaction. The high-energy photon flux accelerator proposed for transmutation of {sup 90}Sr and {sup 137}Cs consumes about an order of magnitude less power than those proposed by Los Alamos National Laboratory (LANL) for the transmutation of nuclear waste in an intense thermal neutron source. Furthermore, this system can be linked to one of the proposed systems and improve the overall transmutation performance of the fission product, especially of these two problematic isotopes.

Transactions, Jun 1, 2008

Transactions of the American Nuclear Society, Dec 31, 1993

ABSTRACT Explosives concealed in trucks or large containers can be detected utilizing a system ba... more ABSTRACT Explosives concealed in trucks or large containers can be detected utilizing a system based on pulse fast neutron analysis (PFNA) or thermal neutron analysis (TNA). These systems are able to determine the spatial distribution of the various elements in interrogated volume. In the design of the above systems, the charged and neutral particles are traced from the source through their arrival time in the detectors. On-line analysis of the signals from the detectors is used to identify the materials which constitute the sample employing statistical and inverse methods. An extensive research program to develop the computational capability to model this process is underway. The results will produce an optimized and cost effective design of a TNA and PFNA system.

Nuclear Science and Engineering, 1982

An analysis of spectral effects that arise from solving the k-, a-, "1-, and f)-eigenvalue formul... more An analysis of spectral effects that arise from solving the k-, a-, "1-, and f)-eigenvalue formulations of the neutron transport equation is presented. Hierarchies of neutron spectra softness are established and expressed in terms of spatial-dependent local indices that are defined for both the core and the reflector of nuclear system configurations. Conclusions regarding the general behavior of the spectrum-dependent integral spectral indices and initial conversion ratios given by the k-, a-, "1-, and f)-eigenvalue equations are also presented. Spectral effects in the core and in the reflector are distinguished by defining separate integral spectral indices for the core and for the reflector. It is shown that the relationship between the spectra given by the k-, a-, "1-, and f)-eigenvalue equations and the spectrum in a corresponding critical configuration depends on the specific physical process that causes deviation from criticality. Nevertheless, some general recommendations are offered regarding the use of a particular eigenvalue equation for specific applications. All conclusions are supported by numerical experiments performed for an idealized thermal system.

Journal of Physics G, Apr 11, 2007

This paper presents the factors that affect the clean room shielding design at the deep-undergrou... more This paper presents the factors that affect the clean room shielding design at the deep-underground science and engineering laboratory (DUSEL), for enhancing detection of weakly interacting massive particles (WIMP), neutrino or double beta decay signals. The goal of the shielded clean room is to minimize the neutron and gamma-ray background and spurious signals, including secondary particles generated within shielding materials, permitting increased probability of detection of signals arising from WIMP and other particles of interest. Various arrangement options of shielded blocks made of paraffin and lead are examined to reduce the background from intrinsic uranium/thorium trace sources and high energy charged cosmic particles that interact with passive shielding material. A new shield block material is proposed and compared with other ordinary shield block arrangements currently used in detector passive shielding designs.

Nuclear Engineering and Design, 2011

The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and deplete... more The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and depleted uranium in separate TRISO particles, in a single fuel rod within a graphite matrix. The TRISO particle volume packing fraction (PF) in the fuel rods is 29%, of which the LEU particle PF is 62%. The lifetime between refuelings is about 476 effective full power days (EFPD). In this paper we assess the possibility of replacing the depleted uranium TRISO particles with thorium TRISO particles, and evaluate the impact of such replacement on fuel cycle length. A preliminary scoping study was performed to determine the most promising fuel rod/zoning configurations. The scoping study indicates that there is advantage to separating the thorium TRISO particles from the LEU particles at the fuel rod level instead of mixing them within a single rod. An axial checkerboard distribution of the fuel rods where all uranium and all thorium rods are interchangeable along the axial direction within the graphite block is the most promising configuration that was identified in this study and can be lead to a fuel cycle length extension of 50-80% relative to the current design, with only a modest increase in the fissile material loading (15-20%). To this advantage can be added the benefit of a significant reduction in nuclear waste and in health risk. This study also lays the foundation for improving the fuel rod arrangement within the graphite block and the graphite blocks within the entire reactor core. The analysis is limited to a once-through fuel cycle based on in situ fissioning of the U-233, without further separation and reprocessing. The preliminary heat transfer analysis indicates that the maximum temperature in the fuel will be raised by about 10-15% over that of current MHR design.

Nuclear Technology, Feb 1, 1988

The view has long been held that breeding in light water cores is possible only with the thorium ... more The view has long been held that breeding in light water cores is possible only with the thorium cycle, at a rate slightly above the break-even point. If we utilize the uranium-plutonium cycle (plutonium fuel with m U fertile material), we find that in a typical light water spectrum the value of r; (number of neutrons emitted per neutron absorbed) for 239 Pu, the principal plutonium isotope in standard light water reactor (LWR) spent fuel, is <2, which is the minimum value necessary for break-even in fissile fuel content. The reason for the low yield of fission neutrons from 239 Pu absorption is that nearly one-third of the time 239 Pu does not fission but instead forms 240 Pu, a relatively nonfissile isotope. However, in studying the effect of neutron absorption by 240 Pu and then by the subsequent 241 Pu, which is highly fissionable with a large value of rj, it is found that a net neutron gain is obtained. This, combined with the neutrons obtained from 239 Pu fission and the fast effect (238 U fissions), yields sufficient neutrons for a high gain breeding potential.

Ceramic transactions, Jan 5, 2018

The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and deplete... more The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and depleted uranium in separate TRISO particles, in a single fuel rod within a graphite matrix. The TRISO particle volume packing fraction (PF) in the fuel rods is 29%, of which the LEU particle PF is 62%. The lifetime between refuelings is about 476 effective full power days (EFPD). In this paper we assess the possibility of replacing the depleted uranium TRISO particles with thorium TRISO particles, and evaluate the impact of such replacement on fuel cycle length. A preliminary scoping study was performed to determine the most promising fuel rod/zoning configurations. The scoping study indicates that there is advantage to separating the thorium TRISO particles from the LEU particles at the fuel rod level instead of mixing them within a single rod. An axial checkerboard distribution of the fuel rods where all uranium and all thorium rods are interchangeable along the axial direction within the graphite block is the most promising configuration that was identified in this study and can be lead to a fuel cycle length extension of 50% to 80% relative to the current design, with only a modest increase in the fissile material loading (15%–20%). To this advantage can be added the benefit of a significant reduction in nuclear waste and in health risk. This study also lays the foundation for improving the fuel rod arrangement within the graphite block and the graphite blocks within the entire reactor core. The analysis is limited to a once-through fuel cycle based on in-situ fissioning of the U-233, without further separation and reprocessing.

Nuclear Science and Engineering, Feb 1, 1987

Since 1963, the INEL has calcined almost 8 million gallons of liquid mixed waste and liquid high-... more Since 1963, the INEL has calcined almost 8 million gallons of liquid mixed waste and liquid high-level waste, converting it to some 1.1 million gallons of dry calcine (about 4275.0 m3), which consists of alumina-and zirconia-based calcine and zirconia-sodium blend calcine. In addition, if all existing and projected future liquid wastes are solidified, approximately 2,000 m3 of additional calcine will be produced primarily from sodium-bearing waste. Calcine is a more desirable material to store than liquid radioactive waste because it reduces volume, is much less corrosive, less chemically reactive, less mobile under most conditions, easier to monitor and more protective of human health and the environment. This paper describes the technical issue involved in the development of a feasible solution for further volume reduction of calcined nuclear waste for transportation and long term storage, using a standard DWPF canister. This will be accomplished by developing a process wherein the canisters are transported into a vibrational machine, for further volume reduction by about 35%. The random compaction experiments show that this volume reduction is achievable. The main goal of this paper is to demonstrate through computer modeling that it is feasible to use volume reduction vibrational machine without developing stress/strain forces that will weaken the canister integrity. Specifically, the paper presents preliminary results of the stress/strain analysis of the DWPF canister as a function of granular calcined height during the compaction and verifying that the integrity of the canister is not compromised. This preliminary study will lead to the development of better technology for safe compactions of nuclear waste that will have significant economical impact on nuclear waste storage and treatment. The preliminary results will guide us to find better solutions to the following questions: 1) What are the optimum locations and directions (vertical versus horizontal or both of them) of applying the vibrational forces? 2) What is best mode of operation: first fill the canister with calcined waste and then vibrate it and refill it again, or apply vibrational forces during the filling process. By optimum or best we mean less creation of stress/strain forces during the volume reduction vibration process. Lessons learnt: This preliminary study shows that; 1) The maximum stress concentration always occurs in the canister wall, however its location varies and depends on the loading condition, and vibration process. 2) The proposed vibrational process would not cause any damages to the granulated calcined waste. 3) The first natural frequency of the longitudinal vibration of the canister is around 400 Hz, which is far away from the applied vibrational frequencies and from possibility of resonance phenomena that may cause damage to the canister 4) The relationship between the maximum internal stress and the frequency of the applied load is not parabolic. 5) The mechanical properties of the granulated calcined nuclear waste have small impact on the internal stress of the canister. Finally, the calculated data suggested that applying vibrational forces will keep the entire canister whole without any indication of development defects, and will have significant economical benefits of handling HLW and LLW in calcined forms, from waste manipulation, storage and transportation.

Transactions of the American Nuclear Society, 1993

Nuclear Engineering and Design, 2017

h i g h l i g h t s We evaluate transmutation fuels for plutonium and minor actinide destruction ... more h i g h l i g h t s We evaluate transmutation fuels for plutonium and minor actinide destruction in LWRs. We model a modified AP1000 fuel assembly in SCALE6.1. We evaluate spectral shift absorber coatings to improve transmutation performance.

Journal of materials engineering and performance, Jun 14, 2024

Nuclear Science and Engineering, 1987

Transactions of the American Nuclear Society, Dec 31, 1994

Transactions, Jun 1, 2008

Transactions of the American Nuclear Society, Dec 31, 1993

Nuclear Science and Engineering, 1982

Journal of Physics G, Apr 11, 2007

Nuclear Engineering and Design, 2011

Nuclear Technology, Feb 1, 1988

Ceramic transactions, Jan 5, 2018

The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and deplete... more The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and depleted uranium in separate TRISO particles, in a single fuel rod within a graphite matrix. The TRISO particle volume packing fraction (PF) in the fuel rods is 29%, of which the LEU particle PF is 62%. The lifetime between refuelings is about 476 effective full power days (EFPD). In this paper we assess the possibility of replacing the depleted uranium TRISO particles with thorium TRISO particles, and evaluate the impact of such replacement on fuel cycle length. A preliminary scoping study was performed to determine the most promising fuel rod/zoning configurations. The scoping study indicates that there is advantage to separating the thorium TRISO particles from the LEU particles at the fuel rod level instead of mixing them within a single rod. An axial checkerboard distribution of the fuel rods where all uranium and all thorium rods are interchangeable along the axial direction within the graphite block is the most promising configuration that was identified in this study and can be lead to a fuel cycle length extension of 50% to 80% relative to the current design, with only a modest increase in the fissile material loading (15%–20%). To this advantage can be added the benefit of a significant reduction in nuclear waste and in health risk. This study also lays the foundation for improving the fuel rod arrangement within the graphite block and the graphite blocks within the entire reactor core. The analysis is limited to a once-through fuel cycle based on in-situ fissioning of the U-233, without further separation and reprocessing.

Nuclear Science and Engineering, Feb 1, 1987

Transactions of the American Nuclear Society, 1993

Nuclear Engineering and Design, 2017