Fission blanket benchmark experiment on spherical assembly of uranium and PE with PE reflector

Energy Technology Data Exchange (ETDEWEB)

Zhu, Tonghua; Lu, Xinxin; Wang, Mei; Han, Zijie, E-mail: neutron_integral@aliyun.com; Jiang, Li; Wen, Zhongwei; Liu, Rong

2016-04-15

Highlights: • The fission rate distribution on two depleted uranium assemblies was measured with plate fission chambers. • We do calculations using MCNP code and ENDF/B-V.0 library. • The overestimation of calculations to the measured fission rates was found. • The observed discrepancy are discussed. - Abstract: New concept of fusion-fission hybrid for energy generation has been proposed. To validate the nuclear performance of fission blanket of hybrid, as part of series of validation experiment, two types of fission blanket assemblies were setup in this work and measurements were made of the reaction rate distribution for uranium fission in the spherical assembly of depleted uranium and polyethylene by Plate Fission Chamber (PFC). There are two PFCs in experiment, one is depleted uranium chamber and the other is enriched uranium chamber. The Monte-Carlo transport code MCNP5 and continuous energy cross sections library ENDF/BV.0 were used for the analysis of fission rate distribution in the two types of assemblies. The calculated results were compared with the experimental ones. The overestimation of fission rate for depleted uranium and enriched uranium were found in the inner boundary of the two assemblies. However, the C/E ratio tends to decrease for the distance from the core slightly and the results for enriched uranium are better than that for depleted uranium.

Fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Greenspan, E.

1984-01-01

This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

The LOFA analysis of fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Yu, Z.-C.; Xie, H.

2014-01-01

The fusion-fission hybrid energy reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc, with the fusion neutron source striking the subcritical blanket. The passive safety system, consisting of passive residual heat removal system, passive safety injection system and automatic depressurization system, was adopted into the fusion-fission hybrid energy reactor in this paper. Modeling and nodalization of primary loop, passive core cooling system and partial secondary loop of the fusion-fission hybrid energy reactor using RELAP5 were conducted and LOFA (Loss of Flow Accident) was analyzed. The results of key transient parameters indicated that the PRHRs could mitigate the accidental consequence of LOFA effectively. It is also concluded that it is feasible to apply the passive safety system concept to fusion-fission hybrid energy reactor. (author)

Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

Science.gov (United States)

Shi, Xue-Ming; Peng, Xian-Jue

2016-09-01

Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

Journey from discovery of nuclear fission to accelerator-driven sub-critical reactor systems (ADS)

International Nuclear Information System (INIS)

Kapoor, S.S.

2005-01-01

The epoch making discovery of nuclear fission in 1939, which resulted purely from the curiosity driven basic research to understand the atomic and nuclear structure has changed the world forever with the onset of a new era in the history of human civilization. The basic nuclear physics research pursued after the discovery of fission has also been of much relevance in the harnessing of nuclear energy. In the recent years, there is considerable interest towards developing accelerator driven sub-critical reactor systems (ADS) for the incineration of the long-lived spent fuel radioactive waste and for the utilization of thorium fuel for nuclear power generation. In this talk, we discuss important milestones in the journey from discovery of nuclear fission to ADS. (author)

Fusion-driven sub-critical dual-cooled waste transmutation blanket: design and analysis

International Nuclear Information System (INIS)

Wang Weihua; Wu Yican; Ke Yan; Kang Zhicheng; Wang Hongyan; Huang Qunying

2003-01-01

The Fusion-Driven Sub-critical System (FDS) is one of the Chinese programs to be further developed for fusion application. Its Dual-cooled Waste Transmutation Blanket (DWTB), as one the most important part of the FDS is cooled by helium and liquid metal, and have the features of safety, tritium self-sustaining, high efficiency and feasibility. Its conceptual design has been finished. This paper is mainly involved with the basic structure design and thermal-hydraulics analysis of DWTB. On the basis of a three-dimensional (3-D) model of radial-toroidal sections of the segment box, thermal temperature gradients and structure analysis made with a comprehensive finite element method (FEM) have been performed with the computer code ANSYS5.7 and computational fluid dynamic finite element codes. The analysis refers to the steady-state operating condition of an outboard blanket segment. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions have been also taken into account. All the above loads have been combined as an input for a FEM stress analysis and the resulting stress distribution has been evaluated. Finally, the structure design and Pb-17Li flow velocity has been optimized according to the calculations and analysis

On the safety of conceptual fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Okrent, D.; Badham, V.; Caspi, S.; Chan, C.K.; Ferrell, W.J.; Frederking, T.H.K.; Grzesik, J.; Lee, J.Y.; McKone, T.E.; Pomraning, G.C.; Ullman, A.Z.; Ting, T.D.; Kim, Y.I.

1979-01-01

A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium. As a result of these studies, it appears that highly reliable and even redundent decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered. (Auth.)

Neutronic design and analysis on dual-cooled waste transmutation blanket for the fusion driven sub-critical system

International Nuclear Information System (INIS)

Zheng Shanliang; Wu Yican; Gao Chunjing; Xu Dezheng; Li Jingjing; Zhu Xiaoxiang

2004-01-01

Neutronics design and analysis of dual-cooled multi-functional waste transmutation blanket (DWTB) for the fusion driven sub-critical system (FDS) are performed to ensure the system be able to meet the requirements of fuel-sufficiency and more waste transmutation ratio with low initial loading fuel inventory, which is based on 1-D burn-up calculations with home-developed code Visual BUS and the multi-group (175 neutron groups-42 Gamma groups coupled) data library HENDL1.0/MG (Hybrid Evaluated Nuclear Data Library). (authors)

Feasibility study of incore fission chamber application for neutron flux measurements on the NET blanket

International Nuclear Information System (INIS)

Bertalot, L.

1987-01-01

A feasibility study has been carried out on the use of in-core fission chambers as neutron diagnostic tools to perform neutron flux measurements on the blanket component of NET. The high neutron and gamma fluxes and the severe thermal-mechanical and magnetic conditions of the blanket structure have been taken into account in this analysis. Preliminary design criteria and specifications of an in-core detector are presented for NET application. A research and development programme is outlined which aims to obtain more information on the tecnological constraints arising from the severe conditions of the NET blanket

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

International Nuclear Information System (INIS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-01-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

Calculation of neutron interior source distribution within subcritical fission-chain reacting systems for a prescribed power density generation

International Nuclear Information System (INIS)

Moraes, Leonardo R.C.; Alves Filho, Hermes; Barros, Ricardo C.

2017-01-01

Accelerator Driven Systems (ADS) are sub-critical systems stabilized by stationary external sources of neutrons. A system is subcritical when the removal by absorption and leakage exceeds the production by fission and tends to shut down. On the other hand, any subcritical system can be stabilized by including time-independent external sources of neutrons. The goal of this work is to determine the intensity of uniform and isotropic sources of neutrons that must be added inside all fuel regions of a subcritical system so that it becomes stabilized, generating a prescribed distribution of electric power. A computer program has been developed in Java language to estimate the intensity of stationary sources of neutrons that must be included in the fuel regions to drive the subcritical system with a fixed power distribution prescribed by the user. The mathematical model used to achieve this goal was the energy multigroup, slab-geometry neutron transport equation in the discrete ordinates (S N ) formulation and the response matrix method was applied to solve the forward and the adjoint S N problems. Numerical results are given to verify the present. (author)

Calculation of neutron interior source distribution within subcritical fission-chain reacting systems for a prescribed power density generation

Energy Technology Data Exchange (ETDEWEB)

Moraes, Leonardo R.C.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: lrcmoraes@iprj.uerj.br, E-mail: halves@iprj.uerj.br, E-mail: ricardob@iprj.uerj.br [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Programa de Pós-Graduação em Modelagem Computacional

2017-07-01

Accelerator Driven Systems (ADS) are sub-critical systems stabilized by stationary external sources of neutrons. A system is subcritical when the removal by absorption and leakage exceeds the production by fission and tends to shut down. On the other hand, any subcritical system can be stabilized by including time-independent external sources of neutrons. The goal of this work is to determine the intensity of uniform and isotropic sources of neutrons that must be added inside all fuel regions of a subcritical system so that it becomes stabilized, generating a prescribed distribution of electric power. A computer program has been developed in Java language to estimate the intensity of stationary sources of neutrons that must be included in the fuel regions to drive the subcritical system with a fixed power distribution prescribed by the user. The mathematical model used to achieve this goal was the energy multigroup, slab-geometry neutron transport equation in the discrete ordinates (S{sub N}) formulation and the response matrix method was applied to solve the forward and the adjoint S{sub N} problems. Numerical results are given to verify the present. (author)

First wall material damage induced by fusion-fission neutron environment

Energy Technology Data Exchange (ETDEWEB)

Khripunov, Vladimir, E-mail: Khripunov_VI@nrcki.ru

2016-11-01

Highlights: • The highest damage and gas production rates are experienced within the first wall materials of a hybrid fusion-fission system. • About ∼2 times higher dpa and 4–5 higher He appm are expected compared to the values distinctive for a pure fusion system at the same DT-neutron wall loading. • The specific nuclear heating may be increased by a factor of ∼8–9 due to fusion and fission neutrons radiation capture in metal components of the first wall. - Abstract: Neutronic performance and inventory analyses were conducted to quantify the damage and gas production rates in candidate materials when used in a fusion-fission hybrid system first wall (FW). The structural materials considered are austenitic SS, Cu-alloy and V- alloys. Plasma facing materials included Be, and CFC composite and W. It is shown that the highest damage rates and gas particles production in materials are experienced within the FW region of a hybrid similar to a pure fusion system. They are greatly influenced by a combined neutron energy spectrum formed by the two-component fusion-fission neutron source in front of the FW and in a subcritical fission blanket behind. These characteristics are non-linear functions of the fission neutron source intensity. Atomic displacement damage production rate in the FW materials of a subcritical system (at the safe subcriticality limit of ∼0.95 and the neutron multiplication factor of ∼20) is almost ∼2 times higher compared to the values distinctive for a pure fusion system at the same 14 MeV neutron FW loading. Both hydrogen (H) and helium (He) gas production rates are practically on the same level except of about ∼4–5 times higher He-production in austenitic and reduced activation ferritic martensitic steels. A proper simulation of the damage environment in hybrid systems is required to evaluate the expected material performance and the structural component residence times.

Research of the Mass Spectra of the Fission Products and Yields of (n, gamma) and (n, 2n) Reactions in a Model Subcritical Uranium Blanket of the Electronuclear System "Energy Plus Transmutation" on Proton Beam of the Dubna Synchrophasotron at 1.5 Ge

CERN Document Server

Chultem, D; Krivopustov, M I; Gerbish, S; Tumendemberel, B; Pavlyuk, A B; Zaveryukha, O S

2002-01-01

This paper is devoted to the research of the spatial distributions of the yields of (n, f), (n, gamma) and (n, 2n) reactions in a two-section model of the uranium blanket electronuclear installation constructed at the Laboratory of High Energies, JINR (Dubna) for experiments according to the program "Research of physical aspects of the electronuclear method of energy production and of radioactive waste transmutation in atomic power-engineering on beams of the synchrophasotron and nuclotron" - project "Energy plus Transmutation". The mass spectrum of the fission products and yields of above reactions in uranium activation detectors placed on the radii of the so-called detector plates is determined. The experimental results testify that the fission of nuclei in the uranium blanket is made by fast neutrons. This conclusion coincides with the result obtained with track integrators of uranium fission.

Neutronics optimization of LiPb-He dual-cooled fuel breeding blanket for the fusion-driven sub-critical system

International Nuclear Information System (INIS)

Zheng Shanliang; Wu Yican

2002-01-01

The concept of the liquid Li 17 Pb 83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR > 1.05) and annual output of 100 kg or more fissile 239 Pu (FBR > 0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimized calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio (Br = Tbr + Fbr) is listed corresponding to different cases

Flibe blanket concept for transmuting transuranic elements and long lived fission products

International Nuclear Information System (INIS)

Gohar, Y.

2000-01-01

A Molten salt (Flibe) fusion blanket concept has been developed to solve the disposition problems of the spent nuclear fuel and the transuranic elements. This blanket concept can achieve the top rated solution, the complete elimination of the transuranic elements and the long-lived fission products. Small driven fusion devices with low neutron wall loading and low neutron fluence can perform this function. A 344-MW integrated fusion power from D-T plasmas for thirty years with an availability factor of 0.75 can dispose of 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. In addition, the utilization of this blanket concept eliminates the need for a geological repository site, which is a major advantage. This application provides an excellent opportunity to develop and to enhance the public acceptance of the fusion energy for the future. The energy from the transmutation process is utilized to produce revenue. Flibe, lithium-lead eutectic, and liquid lead are possible candidates. The liquid blankets have several features, which are suited for W application. It can operate at constant thermal power without interruption for refueling by adjusting the concentration of the transuranic elements and lithium-6. These liquids operate at low-pressure, which reduces the primary stresses in the structure material. Development and fabrication costs of solid transuranic materials are eliminated. Burnup limit of the transuranic elements due to radiation effects is eliminated. Heat is generated within the liquid, which simplifies the heat removal process without producing thermal stresses. These blanket concepts have large negative temperature coefficient with respect to the blanket reactivity, which enhances the safety performance. These liquids are chemically and thermally stable under irradiation conditions, which minimize the radioactive waste volume. The operational record of the Molten Salt Breeder Reactor with Flibe was very successful

Accelerator driven heavy water blanket on circulating fuel

International Nuclear Information System (INIS)

Kazaritsky, V.D.; Blagovolin, P.P.; Mladov, V.R.; Okhlopkov, M.L.; Batyaev, V.F.; Stepanov, N.V.; Seliverstov, V.V.

1997-01-01

A conceptual design of a heavy water blanket with circulating fuel for an accelerator driven transmutation system is described. The hybrid system consists of a high-current linear accelerator of protons and 4 targets, each placed inside a subcritical blanket

Design and producing of fine-group cross section library HENDL3.0/FG for subcritical system

International Nuclear Information System (INIS)

Zou, J.; Zeng, Q.; Xu, D.; Hu, L.; Long, P.

2012-01-01

To improve the accuracy of the neutron analyses for subcritical system with thermal fission blanket, a coupled neutron and photon (315 n + 42γ) fine-group cross section library HENDL3.0/FG based on ENDF/B-VII, JEFF3.1 and JENDL3.3 was produced by FDS team. In order to test the availability and reliability of the HENDL3.0/FG data library, shielding and critical safety benchmarks were performed with VisualBUS code. The testing results indicated that the discrepancy between calculation and experimental values of nuclear parameters fell in a reasonable range. It showed that the nuclear data library had accuracy and availability. (authors)

Critical and subcritical mass calculations of fissionable nuclides based on JENDL-3.2+

International Nuclear Information System (INIS)

Okuno, H.

2002-01-01

We calculated critical and subcritical masses of 10 fissionable actinides ( 233 U, 235 U, 238 Pu, 239 Pu, 241 Pu, 242m Am, 243 Cm, 244 Cm, 249 Cf and 251 Cf) in metal and in metal-water mixtures (except 238 Pu and 244 Cm). The calculation was made with a combination of a continuous energy Monte Carlo neutron transport code, MCNP-4B2, and the latest released version of the Japanese Evaluated Nuclear Data Library, JENDL-3.2. Other evaluated nuclear data files, ENDF/B-VI, JEF-2.2, and JENDL-3.3 in its preliminary version were also applied to find differences in results originated from different nuclear data files. For the so-called big three fissiles ( 233 U, 235 U and 239 Pu), analyzing the criticality experiments cited in ICSBEP Handbook validated the code-library combination, and calculation errors were consequently evaluated. Estimated critical and lower limit critical masses of the big three in a sphere with/without a water or SS-304 reflector were supplied, and they were compared with the subcritical mass limits of ANS-8.1. (author)

A proton-driven, intense, subcritical, fission neutron source for radioisotope production

Energy Technology Data Exchange (ETDEWEB)

Jongen, Y. [Chemin du Cyclotron, Louvain-la-Neuve (Belgium)

1995-10-01

{sup 99m}Tc, the most frequently used radioisotope in nuclear medicine, is distributed as {sup 99}Mo=>{sup 99m}Tc generators. {sup 99}Mo is a fission product of {sup 235}U. To replace the aging nuclear reactors used today for this production, the author proposes to use a spallation neutron source, with neutron multiplication by fission. A 150 MeV, H{sup {minus}} cyclotron can produce a 225 kW proton beam with 50% total system energy efficiency. The proton beam would hit a molten lead target, surrounded by a water moderator and a graphite reflector, producing around 0.96 primary neutron per proton. The primary spallation neutrons, moderated, would strike secondary targets containing a subcritical amount of {sup 235}U. The assembly would show a k{sub eff} of 0.8, yielding a fivefold neutron multiplication. The thermal neutron flux at the targets location would be 2 {times} 10{sup 14} n/cm{sup 2}.s, resulting in a fission power of 500 to 750 kW. One such system could supply the world demand in {sup 99}Mo, as well as other radioisotopes. Preliminary indications show that the cost would be lower than the cost of a commercial 10 MW isotope production reactor. The cost of operation, of disposal of radiowaste and of decommissioning should be significantly lower as well. Finally, the non-critical nature of the system would make it more acceptable for the public than a nuclear reactor and should simplify the licensing process.

Preconceptual design and analysis of a solid-breeder blanket test in an existing fission reactor

International Nuclear Information System (INIS)

Deis, G.A.; Hsu, P.Y.; Watts, K.D.

1983-01-01

Preconceptual design and analysis have been performed to examine the capabilities of a proposed fission-based test of a water-cooled Li 2 O blanket concept. The mechanical configuration of the test piece is designed to simulate a unit cell of a breeder-outside-tube concept. This test piece will be placed in a fission test reactor, which provides an environment similar to that in a fusion reactor. The neutron/gamma flux from the reactor produces prototypical power density, tritium production rates, and operating temperatures and stresses. Steady-state tritium recovery from the test piece can be attained in short-duration (5-to-6-day) tests. The capabilities of this test indicate that fission-based testing can provide important near-term engineering information to support the development of fusion technology

Conceptual design of the blanket and power conversion system for a mirror hybrid fusion-fission reactor. 12-month progress report, July 1, 1975--June 30, 1976

International Nuclear Information System (INIS)

Schultz, K.R.; Baxi, C.B.; Rao, R.

1976-01-01

This report presents the conceptual design and preliminary feasibility assessment for the hybrid blanket and power conversion system of the Mirror Hybrid Fusion-Fission Reactor. Existing gas-cooled fission reactor technology is directly applicable to the Mirror Hybrid Reactor. There are a number of aspects of the present conceptual design that require further design and analysis effort. The blanket and power conversion system operating parameters have not been optimized. The method of supporting the blanket modules and the interface between these modules and the primary loop helium ducting will require further design work. The means of support and containment of the primary loop components must be studied. Nevertheless, in general, the conceptual design appears quite feasible

Probabilistic safety assessment of the dual-cooled waste transmutation blanket for the FDS-I

International Nuclear Information System (INIS)

Hu, L.; Wu, Y.

2006-01-01

The subcritical dual-cooled waste transmutation (DWT) blanket is one of the key components of fusion-driven subcritical system (FDS-I). The probabilistic safety assessment (PSA) can provide valuable information on safety characteristics of FDS-I to give recommendations for the optimization of the blanket concepts and the improvement of the design. Event tree method has been adopted to probabilistically analyze the safety of the DWT blanket for FDS-I using the home-developed PSA code RiskA. The blanket melting frequency has been calculated and compared with the core melting frequencies of PWRs and a fast reactor. Sensitivity analysis of the safety systems has been performed. The results show that the current preliminary design of the FDS-I is very attractive in safety

Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

International Nuclear Information System (INIS)

Yu, Zhang-cheng; Xie, Heng

2014-01-01

The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs

Subcritical nuclear assembly

Energy Technology Data Exchange (ETDEWEB)

Vega C, H. R., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

2014-08-15

A Subcritical Nuclear Assembly is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the Wigner-Seitz method in the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. This reactor has approximately 2500 kg of natural uranium heterogeneously distributed in slugs. The reactor uses a {sup 239}PuBe neutron source that is located in the center of an hexagonal array. Using Monte Carlo methods, with the MCNP5 code, a three-dimensional model of the subcritical reactor was designed to estimate the effective multiplication factor, the neutron spectra, the total and thermal neutron fluences along the radial and axial axis. With the neutron spectra in two locations outside the reactor the ambient dose equivalent were estimated. (Author)

Subcritical nuclear assembly

International Nuclear Information System (INIS)

Vega C, H. R.

2014-08-01

A Subcritical Nuclear Assembly is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the Wigner-Seitz method in the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. This reactor has approximately 2500 kg of natural uranium heterogeneously distributed in slugs. The reactor uses a 239 PuBe neutron source that is located in the center of an hexagonal array. Using Monte Carlo methods, with the MCNP5 code, a three-dimensional model of the subcritical reactor was designed to estimate the effective multiplication factor, the neutron spectra, the total and thermal neutron fluences along the radial and axial axis. With the neutron spectra in two locations outside the reactor the ambient dose equivalent were estimated. (Author)

Discriminators for the Accelerator-Based Conversion (ABC) concept using a subcritical molten salt system

International Nuclear Information System (INIS)

Arthur, E.; Busksa, J.; Davidson, W.; Poston, D.

1995-05-01

Discriminators are described that quantify enhancements added to plutonium destruction and/or nuclear waste transmutation systems through use of an accelerator/fluid fuel combination. This combination produces a robust and flexible nuclear system capable of the destruction of all major long-lived actinides (including plutonium) and fission products. The discriminators discussed in this report are (1) impact of subcritical operation on safety, (2) impact of subcritical and fluid fuel operation on plutonium burnout scenarios, and (3) neutron economy enhancements brought about by subcritical operation. Neutron economy enhancements are quantified through assessment of long-term dose reduction resulting from transmutation of key fission products along with relaxation of processing frequencies afforded by subcritical operation

Development of fine-group (315n/42γ) cross section library ENDL3.0/FG for fusion-fission hybrid systems

International Nuclear Information System (INIS)

Zeng Qin; Zou Jun; Xu Dezhen; Jiang Jieqiong; Wang Minghuang; Wu Yican; Qiu Yuefeng; Chen Zhong; Chen Yan

2011-01-01

To improve the accuracy of the neutron analyses for subcritical systems with thermal fission blanket, a coupled neutron and photon (315 n + 42γ) fine-group cross section library HENDL3.0/FG based on ENDF/B-Ⅶ. 0 has been produced by FDS team. In order to test the availability and reliability of the HENDL3.0/FG data library, shielding and critical safety benchmarks were performed with VisualBUS code. The testing results indicated that the discrepancy between calculation and experimental values of nuclear parameters fell in a reasonable range. (authors)

A new approach to make collapsed cross section for burnup calculation of subcritical system

International Nuclear Information System (INIS)

Matsunaka, Masayuki; Kondo, Keitaro; Miyamaru, Hiroyuki; Murata, Isao

2008-01-01

A general-purpose transport and burnup code system for precise analysis of subcritical reactors like a fusion-fission (FF) hybrid reactor was developed and used for analyzing their performance. The FF hybrid reactor is a subcritical system, which has a concept of fusion reactor with a blanket region containing nuclear fuel and has been under discussion by author's group for years because the present burnup calculation system mainly consists of a general-purpose Monte Carlo code MCNP-4B, a point burnup code ORIGEN2. JENDL-3.3 pointwise cross section library and JENDL Activation Cross Section File 96 were used as base cross section libraries to make group constant for burnup calculation. A new method has been proposed to make group constant for the burnup calculation as accurate as possible directly using output data of the neutron transport calculation by MCNP and evaluated nuclear data libraries. This method is strict and a general procedure to make one group cross sections in Monte Carlo calculations, while it takes very long computation time. Some speed-up techniques were discussed for the present group constant making process so as to decrease calculation time. Adoption of postprocessing to make group constant improved the calculation accuracy because of increasing number of cross sections to be updated in each burnup cycle. The present calculation system is capable of performing neutronics analysis of subcritical reactors more precise than our previous one. However, at the moment, it still takes long computation time to make group constants. Further speed-up techniques are now under investigation so as to apply the present system to neutronics design analysis for various subcritical systems. (author)

Conceptual research on reactor core physics for accelerator driven sub-critical reactor

International Nuclear Information System (INIS)

Zhao Zhixiang; Ding Dazhao; Liu Guisheng; Fan Sheng; Shen Qingbiao; Zhang Baocheng; Tian Ye

2000-01-01

The main properties of reactor core physics are analysed for accelerator driven sub-critical reactor. These properties include the breeding of fission nuclides, the condition of equilibrium, the accumulation of long-lived radioactive wastes, the effect from poison of fission products, as well as the thermal power output and the energy gain for sub-critical reactor. The comparison between thermal and fast system for main properties are carried out. The properties for a thermal-fast coupled system are also analysed

ITEP Subcritical Neutron Generator driven by charged particle accelerator

Energy Technology Data Exchange (ETDEWEB)

Shvedov, O.V.; Chuvilo, I.V.; Vasiliev, V.V. [Institute of Theoretical and Experimental Physics, Moscow (Russian Federation)] [and others

1995-10-01

A research facility prototype including a combination of a linear accelerator, a neutron generating target, a nuclear safety ensuring and means of its attainment for Subcritical Neutron Generator are considered. The scheme of the multiplying is shown. The assembly will be mounted in the body of the partly dismantled ITEP HWR. Requirements for subcritical assembly are worked out and their feasibility within the framework of the heavy-water blanket is shown. The facility`s application as a full-scale model of more powerful installations of this kind and for fundamental experimental research has been investigated.

Investigation of spatial distribution of fission-rate of natural uranium nuclei in the blanket of electronuclear setup 'Energy plus Transmutation' at Dubna Nuclotron proton beam at energy 1.5 GeV

International Nuclear Information System (INIS)

Hashemi-Nezhad, S.R.; Zhuk, I.V.; Kievets, M.; Krivopustov, M.I.; Sosnin, A.N.; Westmeier, W.; Brandt, R.

2008-01-01

The 'Energy plus Transmutation' experimental setup of the Veksler and Baldin Laboratory of High Energy Physics within the Joint Institute for Nuclear Research (JINR) in Dubna, Russia, is a lead target (with a diameter of 8.4 cm and length of 45.6 cm) surrounded by a uranium blanket (weight 206.4 kg of natural uranium). A polyethylene plus cadmium shield is placed around the target-blanket assembly to modify the spallation and fission neutron spectra in the system. The setup was irradiated by a proton beam of energy 1.5 GeV using the Nuclotron accelerator. The spatial distribution of natural uranium fission-rate in the assembly and fission-rate in the blanket was determined experimentally and compared with Monte Carlo predictions using the MCNPX 2.6C code. Besides neutron-induced fission the calculations include the Nat U(p,f), Nat U(π,f) as well as Nat U(γ,f) reactions. Good agreement between the experimental and calculation results was obtained. The possible sources of errors in the experiment and calculations are discussed in detail

LIFE Materials: Thermomechanical Effects Volume 5 - Part I

International Nuclear Information System (INIS)

Caro, M.; DeMange, P.; Marian, J.; Caro, A.; Fluss, M.; Zepeda-Ruiz, L.

2009-01-01

Improved fuel performance is a key issue in the current Laser Inertial-Confinement Fusion-Fission Energy (LIFE) engine design. LIFE is a fusion-fission engine composed of a ∼40-tons fuel blanket surrounding a pulsed fusion neutron source. Fusion neutrons get multiplied and moderated in a Beryllium blanket before penetrating the subcritical fission blanket. The fuel in the blanket is composed of millions of fuel pebbles, and can in principle be burned to over 99% FIMA without refueling or reprocessing. This report contains the following chapters: Chapter A: LIFE Requirements for Materials -- LIFE Fuel; Chapter B: Summary of Existing Knowledge; Chapter C: Identification of Gaps in Knowledge and Vulnerabilities; and Chapter D: Strategy and Future Work.

Mirror fusion--fission hybrids

International Nuclear Information System (INIS)

Lee, J.D.

1978-01-01

The fusion-fission concept and the mirror fusion-fission hybrid program are outlined. Magnetic mirror fusion drivers and blankets for hybrid reactors are discussed. Results of system analyses are presented and a reference design is described

Modeling of the CTEx subcritical unit using MCNPX code

International Nuclear Information System (INIS)

Santos, Avelino; Silva, Ademir X. da; Rebello, Wilson F.; Cunha, Victor L. Lassance

2011-01-01

The present work aims at simulating the subcritical unit of Army Technology Center (CTEx) namely ARGUS pile (subcritical uranium-graphite arrangement) by using the computational code MCNPX. Once such modeling is finished, it could be used in k-effective calculations for systems using natural uranium as fuel, for instance. ARGUS is a subcritical assembly which uses reactor-grade graphite as moderator of fission neutrons and metallic uranium fuel rods with aluminum cladding. The pile is driven by an Am-Be spontaneous neutron source. In order to achieve a higher value for k eff , a higher concentration of U235 can be proposed, provided it safely remains below one. (author)

The Radiological and Thermal Characteristics of Fission Waste from a Deep-Burn Fusion-Fission Hybrid (LIFE) and Implications for Repository Performance

International Nuclear Information System (INIS)

Shaw, H.F.; Blink, J.; Farmer, J.; Latkowski, J.; Kramer, K.

2009-01-01

We are studying the use of a Laser Inertial-confinement Fusion Engine (LIFE) to drive a hybrid fusion-fission system that can generate electrical power and/or burn nuclear waste. The system uses the neutrons from laser driven ICF to produce tritium and to drive nuclear reactions in a subcritical fission blanket. The fusion neutron source obviates the need for a self-sustaining chain reaction in the fission blanket. Either fissile or fertile could be used as fission fuel, thus eliminating the need for isotopic enrichment. The 'driven' system potentially allows very high levels of burnup to be reached, extracting a large fraction of the available energy in the fission fuel without the need for reprocessing. In this note, we discuss the radionuclide inventory of a depleted uranium (DU) fuel burned to greater than 95% FIMA (Fissions per Initial heavy Metal Atom), the implications for thermal management of the resulting waste, and the implications of this waste for meeting the dose standards for releases from a geological repository for high-level waste. The fission waste discussed here would be that produced by a LIFE hybrid with a 500-MW fusion source. The fusion neutrons are multiplied and moderated by a sequence of concentric shells of materials before encountering the fission fuel, and fission in this region is largely due to thermal neutrons. The fission blanket consists of 40 metric tons (MT) of DU, assumed to be in the form of TRISO-like UOC fuel particles embedded in 2-cm-diameter graphite pebbles. (It is recognized that TRISO-based fuel may not reach the high burnup of the fertile fuel considered here, and other fuel options are being investigated. We postulate the existence of a fuel that can reach >95% FIMA so that the waste disposal implications of high burnup can be assessed.) The engine and plant design considered here would receive one load of fission fuel and produce ∼2 GWt of power (fusion + fission) over its 50- to 70-year lifetime. Neutron and

Economic analysis of the fusion-driven subcritical system

International Nuclear Information System (INIS)

Huang Desuo; Wu Yican; Chu Delin; Hu Liqin

2004-01-01

The economic performance of the Fusion-Driven Subcritical system (FDS) is discussed. At first, as an example, the impacts of parameters, such as plasma aspect-ratio, elongation, normalized beta, on-axis toroidal field and the blanket energy-gain are analyzed on the costs of the typical case (moderate aspect-ratio) of FDS. Then, the economic characteristics of the 3 possible scenarios of FDS are estimated with respect to the neutronics parameters. The results calculated with the SYSCODE developed by the FDS team show that the cost of electricity of Scenario-1 (low aspect-ratio) and Scenario-2 (moderate aspect-ratio) of FDS is cheaper than that of pure fusion power plant at the same plane size (1 GW e ). The cost of electricity of the FDS power plant depends heavily on the functions of blanket and the blanket energy-gain. (authors)

The Complete Burning of Weapons Grade Plutonium and Highly Enriched Uranium with (Laser Inertial Fusion-Fission Energy) LIFE Engine

Energy Technology Data Exchange (ETDEWEB)

Farmer, J C; Diaz de la Rubia, T; Moses, E

2008-12-23

The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

Neutronic Parametric Study on a Conceptual Design for a Transmutation Fusion Blanket

International Nuclear Information System (INIS)

Tariq Siddique, M.; Kim, Myung Hyun

2011-01-01

Fusion energy may be the one of options of future energy. In all over the world, researchers are putting their efforts for its commercial and economical availability. Fusion-fission hybrid reactors have been studied for various applications in China. First milestone of fusion energy is expected to be the fusion fission hybrid reactors. In fusion-fission hybrid reactor the blanket design is of second prime importance after fusion source. In this study conceptual design of a fusion blanket is initiated for calculation of tritium production, transmutation of minor actinides (MA) and fission products (FP) and energy multiplication calculations

Summary of the target-blanket breakout group

Energy Technology Data Exchange (ETDEWEB)

Capiello, M.; Bell, C. [Los Alamos National Laboratory, NM (United States); Barthold, W.

1995-10-01

This breakout group discussed a number of topics and issues pertaining to target and blanket concepts for accelerator-driven systems. This major component area is one marked by a broad spectrum of technical approaches. It is therefore less defined than other major component areas such as the accelerator and is at an earlier stage of technical needs and task specification. The working group did reach a number of general conclusions and recommendations that are summarized. The Conference and the Target/Blanket Breakout Group provided a first opportunity for people working on a variety of missions and concepts to get together and exchange information. A number of subcritical systems applicable for a spectrum of missions were proposed at the Conference and discussed in the Breakout Group. Missions included plutonium disposition, energy production, waste destruction, isotope production, and neutron scattering. The Target/Blanket Breakout Group also defined areas where parameters and data should be addressed as target/blanket design activities become more detailed and sophisticated.

Fusion-Fission Hybrid for Fissile Fuel Production without Processing

Energy Technology Data Exchange (ETDEWEB)

Fratoni, M; Moir, R W; Kramer, K J; Latkowski, J F; Meier, W R; Powers, J J

2012-01-02

Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of {sup 232}Th and fission of {sup 233}U in situ without reprocessing or 'closed' cycles based on irradiation of {sup 232}Th followed by reprocessing, and recycling of {sup 233}U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE1) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in

Hefei experimental hybrid fusion-fission reactor conceptual design

International Nuclear Information System (INIS)

Qiu Lijian; Luan Guishi; Xu Qiang

1992-03-01

A new concept of hybrid reactor is introduced. It uses JET-like(Joint European Tokamak) device worked at sub-breakeven conditions, as a source of high energy neutrons to induce a blanket fission of depleted uranium. The solid breeding material and helium cooling technique are also used. It can produce 100 kg of 239 Pu per year by partial fission suppressed. The energy self-sustained of the fusion core is not necessary. Plasma temperature is maintained by external 20 MW ICRF (ion cyclotron resonance frequency) and 10 MW ECRF (electron cyclotron resonance frequency) heating. A steady state plasma current at 1.5 Ma is driven by 10 MW LHCD (lower hybrid current driven). Plasma density will be kept by pellet injection. ICRF can produce a high energy tail in ion distribution function and lead to significant enhancement of D-T reaction rate by 2 ∼ 5 times so that the neutron source strength reaches to the level of 1 x 10 19 n/s. This system is a passive system. It's power density is 10 W/cm 3 and the wall loading is 0.6 W/cm 2 that is the lower limitation of fusion and fission technology. From the calculation of neutrons it could always be in sub-critical and has intrinsic safety. The radiation damage and neutron flux distribution on the first wall are also analyzed. According to the conceptual design the application of this type hybrid reactor earlier is feasible

Using anisotropies in prompt fission neutron coincidences to assess the neutron multiplication of highly multiplying subcritical plutonium assemblies

Energy Technology Data Exchange (ETDEWEB)

Mueller, J.M., E-mail: jonathan_mueller@ncsu.edu; Mattingly, J.

2016-07-21

There is a significant and well-known anisotropy between the prompt neutrons emitted from a single fission event; these neutrons are most likely to be observed at angles near 0° or 180° relative to each other. However, the propagation of this anisotropy through different generations of a fission chain reaction has not been previously studied. We have measured this anisotropy in neutron–neutron coincidences from a subcritical highly-multiplying assembly of plutonium metal. The assembly was a 4.5 kg α-phase plutonium metal sphere composed of 94% {sup 239}Pu and 6% {sup 240}Pu by mass. Data were collected using two EJ-309 liquid scintillators and two EJ-299 plastic scintillators. The angular distribution of neutron–neutron coincidences was measured at 90° and 180° and found to be largely isotropic. Simulations were performed using MCNPX-PoliMi of similar plutonium metal spheres of varying sizes and a correlation between the neutron multiplication of the assembly and the anisotropy of neutron–neutron coincidences was observed. In principle, this correlation could be used to assess the neutron multiplication of an unknown assembly.

Materials compatibility considerations for a fusion-fission hybrid reactor design

International Nuclear Information System (INIS)

DeVan, J.H.; Tortorelli, P.F.

1983-01-01

The Tandem Mirror Hybrid Reactor is a fusion reactor concept that incorporates a fission-suppressed breeding blanket for the production of 233 U to be used in conventional fission power reactors. The present paper reports on compatibility considerations related to the blanket design. These considerations include solid-solid interactions and liquid metal corrosion. Potential problems are discussed relative to the reference blanket operating temperature (490 0 C) and the recycling time of breeding materials (<1 year)

Neutronic calculation and cross section sensitivity analysis of the Livermore mirror fusion/fission hybrid reactor blanket

International Nuclear Information System (INIS)

Ku, L.P.; Price, W.G. Jr.

1977-08-01

The neutronic calculation for the Livermore mirror fusion/fission hybrid reactor blanket was performed using the PPPL cross section library. Significant differences were found in the tritium breeding and plutonium production in comparison to the results of the LLL calculation. The cross section sensitivity study for tritium breeding indicates that the response is sensitive to the cross section of 238 U in the neighborhood of 14 MeV and 1 MeV. The response is also sensitive to the cross sections of iron in the vicinity of 14 MeV near the first wall. Neutron transport in the resonance region is not important in this reactor model

Design status and future research programme for a sub-critical assembly driven by a proton accelerator with proton energy 660 MeV for experiments on long-lived fission products and minor actinides transmutation (Sad)

International Nuclear Information System (INIS)

Gustov, S.A.; Mirokhin, I.V.; Morozov, N.A.; Onischenko, L.M.; Savchenko, O.V.; Sissakian, A.N.; Shvetsov, V.N.; Tretyakov, I.T.; Lopatkin, A.V.; Vorontsov, M.T.

2003-01-01

Report presents project for the construction of a low power integral system on the basis of the proton accelerator of energy 660 MeV and sub-critical MOX blanket with uranium-plutonium fuel. Installation includes sub-critical core with a nominal thermal power of 15-20 kW. Multiplication coefficient k eff = 0,95 and the accelerator beam power of 0.75-1 kW. The experimental programme for SAD will be focused on solving different aspects of reactor physics, reaction rates measurements and benchmarking. The first conceptual design of the SAD experiment is completed in the form of the ISTC Project Proposal 2267. Realisation of the SAD facility may be expected in about 3-4 years. (author)

Subcritical neutron generator-test facility for nuclear waste transmutation studies

Energy Technology Data Exchange (ETDEWEB)

Chuvilo, I.V.; Kolomiets, A.A.; Kozodaev, A.M. [ITEP, Moscow (Russian Federation)] [and others

1995-10-01

The development of the optimal design of high power facility for NPP transmutation and for a number of applications can not be carried out without preliminary tests of much cheaper prototypes. It has been proposed to combine in new test facility 36 MeV Linac ISTRA constructed in ITEP, original Be target and subcritical blanket that will be mounted on the place of partly disassembled heavy water ITEP experimental reactor. The basic parameters of Linac, schemes of the target and blanket are described. It will provide the direct experiments on installation which can be considered as prototype for future linac driven high power facilities.

Workshop on Subcritical Neutron Production

International Nuclear Information System (INIS)

Walter Sadowski; Roald Sagdeev

2006-01-01

Executive Summary of the Workshop on Subcritical Neutron Production A workshop on Subcritical Neutron Production was sponsored by the East-West Center of the University of Maryland on October 11-13, 2004. The subject of the workshop was the application of subcritical neutrons to transmutation of actinides. The workshop was attended by members of the fission, accelerator and fusion communities. Papers on the state of development of neutron production by accelerators, fusion devices, and fission reactors were presented. Discussions were held on the potential of these technologies to solve the problems of spent nuclear waste storage and nuclear non-proliferation presented by current and future nuclear power reactors. A list of participants including their affiliation and their E-Mail addresses is attached. The workshop concluded that the technologies, presently available or under development, hold out the exciting possibility of improving the environmental quality and long term energy resources of nuclear power while strengthening proliferation resistance. The workshop participants agreed on the following statements. The workshop considered a number of technologies to deal with spent nuclear fuels and current actinide inventories. The conclusion was reached that substantial increase in nuclear power production will require that the issue of spent nuclear fuel be resolved. The Workshop concluded that 14 MeV fusion neutrons can be used to destroy nuclear reactor by-products, some of which would otherwise have to be stored for geologic periods of time. The production of 14 MeV neutrons is based on existing fusion technologies at different research institutions in several countries around the world. At the present time this technology is used to produce 14 MeV neutrons in JET. More development work will be required, however, to bring fusion technology to the level where it can be used for actinide burning on an industrial scale. The workshop concluded that the potential

First Wall, Blanket, Shield Engineering Technology Program

International Nuclear Information System (INIS)

Nygren, R.E.

1982-01-01

The First Wall/Blanket/Shield Engineering Technology Program sponsored by the Office of Fusion Energy of DOE has the overall objective of providing engineering data that will define performance parameters for nuclear systems in advanced fusion reactors. The program comprises testing and the development of computational tools in four areas: (1) thermomechanical and thermal-hydraulic performance of first-wall component facsimiles with emphasis on surface heat loads; (2) thermomechanical and thermal-hydraulic performance of blanket and shield component facsimiles with emphasis on bulk heating; (3) electromagnetic effects in first wall, blanket, and shield component facsimiles with emphasis on transient field penetration and eddy-current effects; (4) assembly, maintenance and repair with emphasis on remote-handling techniques. This paper will focus on elements 2 and 4 above and, in keeping with the conference participation from both fusion and fission programs, will emphasize potential interfaces between fusion technology and experience in the fission industry

Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

Energy Technology Data Exchange (ETDEWEB)

Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio [Universidad Simón Bolívar, Nuclear Physics Laboratory, Apdo 89000, Caracas 1080A (Venezuela, Bolivarian Republic of); Davila, Jesus [Física Médica C. A. and Universidad Central de Venezuela, Caracas (Venezuela, Bolivarian Republic of)

2015-07-23

During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e’n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides {sup 232}Th, {sup 238}U and {sup 237}Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

Science.gov (United States)

Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

2015-07-01

During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

Burn-up calculation of fusion-fission hybrid reactor using thorium cycle

International Nuclear Information System (INIS)

Shido, S.; Matsunaka, M.; Kondo, K.; Murata, I.; Yamamoto, Y.

2006-01-01

A burn-up calculation system has been developed to estimate performance of blanket in a fusion-fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In this system, neutron flux is calculated by MCNP4B and then burn-up calculation is performed by ORIGEN2. The cross-section library for ORIGEN2 is made from the calculated neutron flux and evaluated nuclear data. The 3-dimensional ITER model was used as a base fusion reactor. The nuclear fuel (reprocessed plutonium as the fission materials mixed with thorium as the fertile materials), transmutation materials (minor actinides and long-lived fission products) and tritium breeder were loaded into the blanket. Performances of gas-cooled and water-cooled blankets were compared with each other. As a result, the proposed reactor can meet the requirement for TBP and power density. As far as nuclear waste incineration is concerned, the gas-cooled blanket has advantages. On the other hand, the water cooled-blanket is suited to energy production. (author)

Concept and nuclear performance of direct-enrichment fusion breeder blanket using UO2 powder

International Nuclear Information System (INIS)

Oka, Yoshiaki; Kasahara, Takayasu; An, Shigehiro

1985-01-01

A new concept is presented for direct enrichment of fissile fuel in the blanket of a fusion-fission hybrid reactor. The enriched fuel produced by this means can be used in fission reactors without reprocessing. The outstanding feature of the concept is the powdered form in which UO 2 fuel is placed in the reactor blanket, where it is irradiated to the requisite enrichment for use as fuel in burner reactor, e.g. 3%. After removal from blanket, the powder is mixed to homogenize the enrichment. Fuel pellets and assemblies are then fabricated from the powder without reprocessing. The concept of irradiating UO 2 in powder eliminates the problems of spatial nonuniformity in fissile enrichment, and of radiation damage to fuel clad, encountered in attempting to enrich prefabricated fuel. Powder mixing for homogenization brings the additional benefit of removing volatile fission products. Also burnable poison can be added, as necessary, after irradiation. An extensive neutronic parameter survey showed that the optimum blanket arrangement for this enrichment concept is one presenting a fission suppressing configuration and with beryllium adopted as moderator. By this arrangement, the average 239 Pu enrichment obtained on the natural UO 2 fuel in the blanket reaches 3% after only 0.56 MW.yr/m"2 exposure. A conceptual design is presented of the blanket, together with associated fusion breeder, from which, practical application of the concept is shown to be promising. (author)

Feasibility of waste transmutation using accelerator-driven IRIS subcritical system

International Nuclear Information System (INIS)

Petroviae, B.; Carelli, M.; Paramonov, D.

2001-01-01

Waste transmutation is considered for reducing radio-toxicity of nuclear waste generated in power reactors. Accelerator driven subcritical systems (ADS) offer certain advantages over the use of nuclear reactors. Transmutation of fission products (e.g. 99 Tc) generally requires thermal neutron spectrum, while for actinides fast spectrum provides better performance. Proposed solutions to this problem include a multi-strata approach as well as a multi-zone (thermal/fast-spectrum) single systems. In this paper we examine the feasibility of employing a dual-spectrum two-zone accelerator-driven IRIS subcritical for waste transmutation. (author)

The spatial kinetic analysis of accelerator-driven subcritical reactor

International Nuclear Information System (INIS)

Takahashi, H.; An, Y.; Chen, X.

1998-02-01

The operation of the accelerator driven reactor with subcritical condition provides a more flexible choice of the reactor materials and of design parameters. A deep subcriticality is chosen sometime from the analysis of point kinetics. When a large reactor is operated in deep subcritical condition by using a localized spallation source, the power distribution has strong spatial dependence, and point kinetics does not provide proper analysis for reactor safety. In order to analyze the spatial and energy dependent kinetic behavior in the subcritical reactor, the authors developed a computation code which is composed of two parts, the first one is for creating the group cross section and the second part solves the multi-group kinetic diffusion equations. The reactor parameters such as the cross section of fission, scattering, and energy transfer among the several energy groups and regions are calculated by using a code modified from the Monte Carlo codes MCNPA and LAHET instead of the usual analytical method of ANISN, TWOTRAN codes. Thus the complicated geometry of the accelerator driven reactor core can be precisely taken into account. The authors analyzed the subcritical minor actinide transmutor studied by Japan Atomic Energy Research Institute (JAERI) using the code

Concept of a subcritical transmutation system with fast neutron spectrum and liquid fuel

International Nuclear Information System (INIS)

Tittelbach, S.

2002-11-01

The annual amount of nearly 9500 t of spent fuel from worldwide industrial nuclear energy utilization has to be disposed as high level waste. The retention of nuclear waste from the biosphere has to be assured until the radiological risk decreases to tolerable levels. The long-term radiological risk of spent fuel is dominated by actinide elements, i.e. plutonium, americium and curium. It is intended to reduce this amount of high level waste by Partitioning and Transmutation, so that the radiotoxicity of the disposed waste falls short of the reference value of fresh fuel decaying naturally after about thousand years. For this time period the retention of high level waste can be assured by technical means. The scope of this work is the design of a subcritical fast transmutation system with liquid metal cooling and liquid metal fuel. The lead bismuth eutectic has been choosen as the liquid metal coolant and fuel carrier. To dissolve at least 3 at% of transuran elements, a minimum fuel temperature of 600 C is required. The calculations were carried out with a fuel composition, which results from two plutonium recycling steps in a thorium fuel cycle. Two homogeneous and two heterogeneous blankets have been designed and evaluated leading to one preferred heterogeneous blanket design, which has been investigated in more detail. This blanket design merges the positive properties of a solid fuel system (better control of fuel and reactivity because of smaller and closed fuel volumina) and a liquid fuel system (continous charge and discharge or extraction of fission products). The blanket design is based on the core design of fast breeder liquid metal reactors. It consists of hexagonal fuel elements housing up to six annular shaped fuel cylinders. The hexagonal shape of the fuel elements leads to three fuel zones positioned concentrically around the central spallation target. There is a strong heterogeneous distribution of power and heat flux in this blanket design. Besides

Candidate molten salt investigation for an accelerator driven subcritical core

International Nuclear Information System (INIS)

Sooby, E.; Baty, A.; Beneš, O.; McIntyre, P.; Pogue, N.; Salanne, M.; Sattarov, A.

2013-01-01

Highlights: • Developing accelerator driven subcritical fission to destroy transuranics in SNF. • The core is a vessel containing a molten mixture of NaCl and transuranic chlorides. • Molecular dynamics used to calculate the thermophysical properties of the salt. • Density and molecular structure for actinide salts reported here. • The neutronics of ADS fission in molten salt are presented. -- Abstract: We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated

Candidate molten salt investigation for an accelerator driven subcritical core

Energy Technology Data Exchange (ETDEWEB)

Sooby, E., E-mail: soobyes@tamu.edu [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Baty, A. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Beneš, O. [European Commission, DG Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); McIntyre, P.; Pogue, N. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Salanne, M. [Université Pierre et Marie Curie, CNRS, Laboratoire PECSA, F-75005 Paris (France); Sattarov, A. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States)

2013-09-15

Highlights: • Developing accelerator driven subcritical fission to destroy transuranics in SNF. • The core is a vessel containing a molten mixture of NaCl and transuranic chlorides. • Molecular dynamics used to calculate the thermophysical properties of the salt. • Density and molecular structure for actinide salts reported here. • The neutronics of ADS fission in molten salt are presented. -- Abstract: We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated.

Neutronics-processing interface analyses for the Accelerator Transmutation of Waste (ATW) aqueous-based blanket system

International Nuclear Information System (INIS)

Davidson, J.W.; Battat, M.E.

1993-01-01

Neutronics-processing interface parameters have large impacts on the neutron economy and transmutation performance of an aqueous-based Accelerator Transmutation of Waste (ATW) system. A detailed assessment of the interdependence of these blanket neutronic and chemical processing parameters has been performed. Neutronic performance analyses require that neutron transport calculations for the ATW blanket systems be fully coupled with the blanket processing and include all neutron absorptions in candidate waste nuclides as well as in fission and transmutation products. The effects of processing rates, flux levels, flux spectra, and external-to-blanket inventories on blanket neutronic performance were determined. In addition, the inventories and isotopics in the various subsystems were also calculated for various actinide and long-lived fission product transmutation strategies

US-DOE Fusion-Breeder Program: blanket design and system performance

International Nuclear Information System (INIS)

Lee, J.D.

1983-01-01

Conceptual design studies are being used to assess the technical and economic feasibility of fusion's potential to produce fissile fuel. A reference design of a fission-suppressed blanket using conventional materials is under development. Theoretically, a fusion breeder that incorporates this fusion-suppressed blanket surrounding a 3000-MW tandem mirror fusion core produces its own tritium plus 5600 kg of 233 U per year. The 233 U could then provide fissile makeup for 21 GWe of light-water reactor (LWR) power using a denatured thorium fuel cycle with full recycle. This is 16 times the net electric power produced by the fusion breeder (1.3 GWe). The cost of electricity from this fusion-fission system is estimated to be only 23% higher than the cost from LWRs that have makeup from U 3 O 8 at present costs (55 $/kg). Nuclear performance, magnetohydrodynamics (MHD), radiation effects, and other issues concerning the fission-suppressed blanket are summarized, as are some of the present and future objectives of the fusion breeder program

Mechanical and thermal design of hybrid blankets

International Nuclear Information System (INIS)

Schultz, K.R.

1978-01-01

The thermal and mechanical aspects of hybrid reactor blanket design considerations are discussed. This paper is intended as a companion to that of J. D. Lee of Lawrence Livermore Laboratory on the nuclear aspects of hybrid reactor blanket design. The major design characteristics of hybrid reactor blankets are discussed with emphasis on the areas of difference between hybrid reactors and standard fusion or fission reactors. Specific examples are used to illustrate the design tradeoffs and choices that must be made in hybrid reactor design. These examples are drawn from the work on the Mirror Hybrid Reactor

Applications of the Aqueous Self-Cooled Blanket concept

International Nuclear Information System (INIS)

Steiner, D.; Embrechts, M.J.; Varsamis, G.; Wrisley, K.; Deutch, L.; Gierszewski, P.

1986-01-01

In this paper a novel water-cooled blanket concept is examined. This concept, designated the Aqueous Self-Cooled Blanket (ASCB), employs water with small amounts of dissolved fertile compounds as both the coolant and the breeding medium. The ASCB concept is reviewed and its application in three different contexts is examined: (1) power reactors; (2) near-term devices such as NET; and (3) fusion-fission hybrids

Advanced burnup calculation code system in a subcritical state with continuous-energy Monte Carlo code for fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Matsunaka, Masayuki; Ohta, Masayuki; Miyamaru, Hiroyuki; Murata, Isao

2009-01-01

The fusion-fission (FF) hybrid reactor is a promising energy source that is thought to act as a bridge between the existing fission reactor and the genuine fusion reactor in the future. The burnup calculation system that aims at precise burnup calculations of a subcritical system was developed for the detailed design of the FF hybrid reactor, and the system consists of MCNP, ORIGEN, and postprocess codes. In the present study, the calculation system was substantially modified to improve the calculation accuracy and at the same time the calculation speed as well. The reaction rate estimation can be carried out accurately with the present system that uses track-length (TL) data in the continuous-energy treatment. As for the speed-up of the reaction rate calculation, a new TL data bunching scheme was developed so that only necessary TL data are used as long as the accuracy of the point-wise nuclear data is conserved. With the present system, an example analysis result for our proposed FF hybrid reactor is described, showing that the computation time could really be saved with the same accuracy as before. (author)

Neutronics of a sub-critical system burning non-recycled LWR waste

International Nuclear Information System (INIS)

Wallenius, J.; Tucek, K.; Gudowski, W.; Sanders, C.

1999-01-01

We have investigated neutronic properties of a subcritical system designed for transmutation of non-recycled TRU discharges from commercial light water reactors. Burnable absorbers (BA) and depleted uranium in variable concentrations are introduced in order to maximize fission to absorption ratios, and to minimize power peaking as well as reactivity losses. The use of nitride fuel raises linear power ratings to 60-110 kW/m while keeping fuel center line temperatures below 1400 K after gap closure. A comparatively small power peaking of 1.5 at a subcriticality level of 0.97 allows for a total core power of 1200 MWth with a corresponding proton beam power of 20 MW at BOL. Core averaged fission to absorption ratios for Np and Am as high as 0.5 are achieved using 10 B enriched B 4 C as BA. Hence, both Pu and minor actinide inventories are reduced during burnup in the here proposed system, mitigating swelling problems arising due to high-activity in MA-based fuels. Disadvantages following BA introduction, such as increase of void coefficients and decrease of negative doppler feedback in conjunction with small values of β eff , are addressed by setting the BOL subcriticality level to 0.97. (author)

Reactor parameters and constants determination by using measurements in subcritical and exponential assembly

International Nuclear Information System (INIS)

Voi, Dante Luiz; Santos Bastos, Wilma dos

1995-01-01

Subcritical and exponential experiments are important for Reactor Physics integral parameter determinations both to validate and confirm theoretical models for reactor calculations. An exponential and subcritical facility has been constructed to be used on the internal thermal column of the Argonauta reactor at IEN-CNEN- Rio de Janeiro. An experimental research program has been developed for the determination of fundamental reactor constants as buckling, migration areas, resonance escape probabilities, thermal utilization, fast fission and fuel eta factors. (author) 23 refs

An optimized symbiotic fusion and molten-salt fission reactor system

International Nuclear Information System (INIS)

Blinkin, V.L.; Novikov, V.M.

A symbiotic fusion-fission reactor system which breeds nuclear fuel is discussed. In the blanket of the controlled thermonuclear reactor (CTR) uranium-233 is generated from thorium, which circulates in the form of ThF 4 mixed with molten sodium and beryllium fluorides. The molten-salt fission reactor (MSR) burns up the uranium-233 and generates tritium for the fusion reactor from lithium, which circulates in the form of LiF mixed with BeF 2 and 233 UF 4 through the MSR core. With a CTR-MSR thermal power ratio of 1:11 the system can produce electrical energy and breed fuel with a doubling time of 4-5 years. The system has the following special features: (1) Fuel reprocessing is much simpler and cheaper than for contemporary fission reactors; reprocessing consists simply in continuous removal of 233 U from the salt circulating in the CTR blanket by the fluorination method and removal of xenon from the MSR fuel salt by gas scavenging; the MSR fuel salt is periodically exchanged for fresh salt and the 233 U is then removed from it; (2) Tritium is produced in the fission reactor, which is a much simpler system than the fusion reactor; (3) The CTR blanket is almost ''clean''; no tritium is produced in it and fission fragment activity does not exceed the activity induced in the structural materials; (4) Almost all the thorium introduced into the CTR blanket can be used for producing 233 U

Method of operating water cooled reactor with blanket

International Nuclear Information System (INIS)

Suzuki, Katsuo.

1988-01-01

Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

Perspective on the fusion-fission energy concept

International Nuclear Information System (INIS)

Liikala, R.C.; Perry, R.T.; Teofilo, V.L.

1978-01-01

A concept which has potential for near-term application in the electric power sector of our energy economy is combining fusion and fission technology. The fusion-fission system, called a hybrid, is distinguished from its pure fusion counterpart by incorporation of fertile materials (uranium or thorium) in the blanket region of a fusion machine. The neutrons produced by the fusion process can be used to generate energy through fission events in the blanket or produce fuel for fission reactors through capture events in the fertile material. The performance requirements of the fusion component of hybrids is perceived as being less stringent than those for pure fusion electric power plants. The performance requirements for the fission component of hybrids is perceived as having been demonstrated or could be demonstrated with a modest investment of research and development funds. This paper presents our insights and observations of this concept in the context of why and where it might fit into the picture of meeting our future energy needs. A bibliography of hybrid research is given

Theoretical and Experimental Research in Neutron Spectra and Nuclear Waste Transmutation on Fast Subcritical Assembly with MOX Fuel

Science.gov (United States)

Arkhipkin, D. A.; Buttsev, V. S.; Chigrinov, S. E.; Kutuev, R. Kh.; Polanski, A.; Rakhno, I. L.; Sissakian, A.; Zulkarneev, R. Ya.; Zulkarneeva, Yu. R.

2003-07-01

The paper deals with theoretical and experimental investigation of transmutation rates for a number of long-lived fission products and minor actinides, as well as with neutron spectra formed in a subcritical assembly driven with the following monodirectional beams: 660-MeV protons and 14-MeV neutrons. In this work, the main objective is the comparison of neutron spectra in the MOX assembly for different external driving sources: a 660-MeV proton accelerator and a 14-MeV neutron generator. The SAD project (JINR, Russia) has being discussed. In the context of this project, a subcritical assembly consisting of a cylindrical lead target surrounded by a cylindrical MOX fuel layer will be constructed. Present conceptual design of the subcritical assembly is based on the core with a nominal unit capacity of 15 kW (thermal). This corresponds to a multiplication coefficient, keff= 0.945, and an accelerator beam power of 0.5 kW. The results of theoretical investigations on the possibility of incinerating long-lived fission products and minor actinides in fast neutron spectrum and formation of neutron spectra with different hardness in subcritical systems based on the MOX subcritical assembly are discussed. Calculated neutron spectra emitted from a lead target irradiated by a 660-MeV protons are also presented.

Transmutation of fission products and actinide waste at Hanford

Energy Technology Data Exchange (ETDEWEB)

Daemen, L.L.; Pitcher, E.J.; Russell, G.J. [Los Alamos National Laboratory, NM (United States)

1995-10-01

The authors studied the neutronics of an ATW system for the transmutation of the fission products ({sup 99}Tc in particular) and the type of actinide waste stored in several tanks at Hanford. The heart of the system is a highly-efficient neutron production target. It is surrounded by a blanket containing a moderator/reflector material, as well as the products to be transmuted. The fission products are injected into the blanket in the form of an aqueous solution in heavy water, whereas an aqueous actinides slurry is circulated in the outer part of the blanket. For the sake of definiteness, the authors focussed on {sup 99}Tc (the most difficult fission product to transmute), and {sup 239}Pu, {sup 237}Np, and {sup 241}Am. Because of the low thermal neutron absorption cross-section of {sup 99}Tc, considerable care and effort must be devoted to the design of a very efficient neutron source.

Enhanced fuel production in thorium/lithium hybrid blankets utilizing uranium multipliers

International Nuclear Information System (INIS)

Pitulski, R.H.

1979-10-01

A consistent neutronics analysis is performed to determine the effectiveness of uranium bearing neutron multiplier zones on increasing the production of U 233 in thorium/lithium blankets for use in a tokamak fusion-fission hybrid reactor. The nuclear performance of these blankets is evaluated as a function of zone thicknesses and exposure by using the coupled transport burnup code ANISN-CINDER-HIC. Various parameters such as U 233 , Pu 239 , and H 3 production rates, the blanket energy multiplication, isotopic composition of the fuels, and neutron leakages into the various zones are evaluated during a 5 year (6 MW.y.m -2 ) exposure period. Although the results of this study were obtained for a tokomak magnetic fusion device, the qualitative behavior associated with the use of the uranium bearing neutron multiplier should be applicable to all fusion-fission hybrids

A novel concept for CRIEC-driven subcritical research reactors

International Nuclear Information System (INIS)

Nieto, M.; Miley, G.H.

2001-01-01

A novel scheme is proposed to drive a low-power subcritical fuel assembly by means of a long Cylindrical Radially-convergent Inertial Electrostatic Confinement (CRIEC) used as a neutron source. The concept is inherently safe in the sense that the fuel assembly remains subcritical at all times. Previous work has been done for the possible implementation of CRIEC as a subcritical assembly driver for power reactors. However, it has been found that the present technology and stage of development of IEC-based neutron sources can not meet the neutron flux requirements to drive a system as big as a power reactor. Nevertheless, smaller systems, such as research and training reactors, could be successfully driven with levels of neutron flux that seem more reasonable to be achieved in the near future by IEC devices. The need for custom-made expensive nuclear fission fuel, as in the case of the TRIGA reactors, is eliminated, and the CRIEC presents substantial advantages with respect to the accelerator-driven subcritical reactors in terms of simplicity and cost. In the present paper, a conceptual design for a research/training CRIEC-driven subcritical assembly is presented, emphasizing the description, principle of operation and performance of the CRIEC neutron source, highlighting its advantages and discussing some key issues that require study for the implementation of this concept. (author)

The safe, economical operation of a slightly subcritical reactor and transmutor with a small proton accelerator

International Nuclear Information System (INIS)

Takahashi, Hiroshi

1994-01-01

I suggest that an accelerator can be used to increase the safety and neutron economy of a power reactor and a transmutor of long-lived radioactive wastes, such as minor actinides and fission products, by providing neutrons for its subcritical operation. Instead of the large subcriticality k=0.9-0.95 which we originally proposed for such transmutor, we propose to use a slightly subcritical reactor, such as k=0.99, which will avoid many of the technical difficulties that are associated with large subcriticality, such as localized power peaking, radiation damage due to injection of medium-energy protons, the high current accelerator, and the requirement for a long beam-expansion section. We analyze the power drop that occurred in Phenix reactor, and show that the operating this reactor in subcritical conditions improves safety. (author). 13 refs., 5 figs

The safe, economical operation of a slightly subcritical reactor and transmutor with a small proton accelerator

International Nuclear Information System (INIS)

Takahashi, Hiroshi.

1994-01-01

This report describes methods in which an accelerator can be used to increase the safety and neutron economy of a power reactor and transmutor of long-lived radioactive wastes, such as minor actinides and fission products, by providing neutrons for its subcritical operation. Instead of the rather large subcriticality of k=0.9--0.95 which we originally proposed for such a transmutor, we propose to use a slightly subcritical reactor, such as k=0.99, which will avoid many of the technical difficulties that are associated with large subcriticality, such as localized power peaking, radiation damage due to the injection of medium-energy protons, the high current accelerator, and the requirement for a long beam-expansion section. We analyzed the power drop that occurred in Phoenix reactor, and show that the operating this reactor in subcritical condition improves its safety

Some new ideas for Tandem Mirror blankets

International Nuclear Information System (INIS)

Neef, W.S. Jr.

1981-01-01

The Tandem Mirror Reactor, with its cylindrical central cell, has led to numerous blanket designs taking advantage of the simple geometry. Also many new applications for fusion neutrons are now being considered. To the pure fusion electricity producers and hybrids producing fissile fuel, we are adding studies of synthetic fuel producers and fission-suppressed hybrids. The three blanket concepts presented are new ideas and should be considered illustrative of the breadth of Livermore's application studies. They are not meant to imply fully analyzed designs

Mirror hybrid reactor blanket and power conversion system conceptual design

International Nuclear Information System (INIS)

Schultz, K.R.; Backus, G.A.; Baxi, C.B.; Dee, J.B.; Estrine, E.A.; Rao, R.; Veca, A.R.

1976-01-01

The conceptual design of the blanket and power conversion system for a gas-cooled mirror hybrid fusion-fission reactor is presented. The designs of the fuel, blanket module and power conversion system are based on existing gas-cooled fission reactor technology that has been developed at General Atomic Company. The uranium silicide fuel is contained in Inconel-clad rods and is cooled by helium gas. The fuel is contained in 16 spherical segment modules which surround the fusion plasma. The hot helium is used to raise steam for a conventional steam cycle turbine generator. The details of the method of support for the massive blanket modules and helium ducts remain to be determined. Nevertheless, the conceptual design appears to be technically feasible with existing gas-cooled technology. A preliminary safety analysis shows that with the development of a satisfactory method of primary coolant circuit containment and support, the hybrid reactor could be licensed under existing Nuclear Regulatory Commission regulations

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

Enhanced fuel production in thorium/lithium hybrid blankets utilizing uranium multipliers

Energy Technology Data Exchange (ETDEWEB)

Pitulski, R.H.

1979-10-01

A consistent neutronics analysis is performed to determine the effectiveness of uranium bearing neutron multiplier zones on increasing the production of U/sup 233/ in thorium/lithium blankets for use in a tokamak fusion-fission hybrid reactor. The nuclear performance of these blankets is evaluated as a function of zone thicknesses and exposure by using the coupled transport burnup code ANISN-CINDER-HIC. Various parameters such as U/sup 233/, Pu/sup 239/, and H/sup 3/ production rates, the blanket energy multiplication, isotopic composition of the fuels, and neutron leakages into the various zones are evaluated during a 5 year (6 MW.y.m/sup -2/) exposure period. Although the results of this study were obtained for a tokomak magnetic fusion device, the qualitative behavior associated with the use of the uranium bearing neutron multiplier should be applicable to all fusion-fission hybrids.

Source-jerk method for application on ADS neutronics study The ADS is stated for Accelerator Driven sub-critical System

CERN Document Server

Zhu Qing Fu; Li Yi; Xia Pu; Zheng Wu Qing; Zhu Guo Sheng

2003-01-01

The paper is concerned in the source-jerk method used to measure the sub-criticality, and the sub-critical experiment facility, which is used for the study on the neutronics of ADS, driven by external neutron source sup 2 sup 5 sup 2 Cf. The effects of the location of neutron source and material buffer where is at the location of the pipe of proton beam and target of fission-product dispersion on the sub-criticality of reactor are studied by source-jerk method

Evaluation of organic moderator/coolants for fusion breeder blankets

International Nuclear Information System (INIS)

Romero, J.B.

1980-03-01

Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process

Conceptual design of a fusion-fission hybrid reactor for transmutation of high level nuclear waste

International Nuclear Information System (INIS)

Qiu, L.J.; Wu, Y.C.; Yang, Y.W.; Wu, Y.; Luan, G.S.; Xu, Q.; Guo, Z.J.; Xiao, B.J.

1994-01-01

To assess the feasibility of the transmutation of long-lived radioactive waste using fusion-fission hybrid reactors, we are studying all the possible types of blanket, including a comparison of the thermal and fast neutron spectrum blankets. Conceptual designs of a small tokamak hybrid blanket with small inventory of actinides and fission products are presented. The small inventory of wastes makes the system safer. The small hybrid reactor system based on a fusion core with experimental parameters to be realized in the near future can effectively transmute actinides and fission products at a neutron wall loading of 1MWm -2 . An innovative energy system is also presented, including a fusion driver, fuel breeder, high level waste transmuter, fission reactor and so on. An optimal combination of all types of reactor is proposed in the system. ((orig.))

Blanket Manufacturing Technologies : Thermomechanical Tests on HCLL Blanket Mocks Up

International Nuclear Information System (INIS)

Laffont, G.; Cachon, L.; Taraud, P.; Challet, F.; Rampal, G.; Salavy, J.F.

2006-01-01

In the Helium Cooled Lithium Lead (HCLL) Blanket concept, the lithium lead plays the double role of breeder and multiplier material, and the helium is used as coolant. The HCCL Blanket Module are made of steel boxes reinforced by stiffening plates. These stiffening plates form cells in which the breeder is slowly flowing. The power deposited in the breeder material is recovered by the breeder cooling units constituted by 5 parallel cooling plates. All the structures such as first wall, stiffening and cooling plates are cooled by helium. Due to the complex geometry of these parts and the high level of pressure and temperature loading, thermo-mechanical phenomena expected in the '' HCLL blanket concept '' have motivated the present study. The aim of this study, carried out in the frame of EFDA Work program, is to validate the manufacturing technologies of HCLL blanket module by testing small scale mock-up under breeder blanket representative operating conditions.The first step of this experimental program is the design and manufacturing of a relevant test section in the DIADEMO facility, which was recently upgraded with an He cooling loop (pressure of 80 bar, maximum temperature of 500 o C,flow rate of 30 g/s) taking the opportunity of synergies with the gas-cooled fission reactor R-and-D program. The second step will deal with the thermo-mechanical tests. This paper focuses on the program made to support the cooling plate mock up tests which will be carried out on the DIADEMO facility (CEA) by thermo-mechanical calculations in order to define the relevant test conditions and the experimental parameters to be monitored. (author)

Advisory group meeting on design and performance of reactor and subcritical blanket systems with lead and lead-bismuth as coolant and/or target material. Working material

International Nuclear Information System (INIS)

2001-01-01

The purpose of the IAEA Advisory Group Meeting (AGM) on Design and Performance of Reactor and Sub-critical Blanket Systems with Lead and Lead-Bismuth as Coolant and/or Target Material was to provide a forum for international information exchange on all the topics relevant to Pb and Pb/Bi cooled critical and sub-critical reactors. In addition, the AGM aimed at: (1) finding ways and means to improve international co-ordination efforts in this area; (2) obtaining advice from the Member States with regard to the activities to be implemented in this area by the IAEA, in order to best meet their needs; and (3) laying out the plans for an effective co-ordination and support of the R and D activities in this area. The AGM stressed that nuclear energy is a realistic solution to satisfy the energy demand, considering the limited resources of fossil fuel, its uneven distribution in the world and the impact of its use on the planet, and taking into account the expected doubling of the world population in the 21st century and tripling of the electricity demand (especially in the developing countries). However, the AGM concluded that the development of an innovative nuclear technology meeting the following requirements must be pursued: (a) deterministic exclusion of any severe accident; (b) proliferation resistance; (c) cost competitiveness with alternative energy sources; (d) sustainable fuel supply; and (e) solution of the radioactive waste management problem

Evaluation of subcritical hybrid systems loaded with reprocessed fuel

International Nuclear Information System (INIS)

Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L.

2015-01-01

Highlights: • Accelerator driven systems (ADS) and fusion–fission systems are investigated for transmutation and fuel regeneration. • The calculations were performed using Monteburns code. • The results indicate the most suitable system for achieve transmutation. - Abstract: Two subcritical hybrid systems containing spent fuel reprocessed by Ganex technique and spiked with thorium were submitted to neutron irradiation of two different sources: ADS (Accelerator-driven subcritical) and Fusion. The aim is to investigate the nuclear fuel evolution using reprocessed fuel and the neutronic parameters under neutron irradiation. The source multiplication factor and fuel depletion for both systems were analysed during 10 years. The simulations were performed using MONTEBURNS code (MCNP/ORIGEN). The results indicate the main differences when irradiating the fuel with different neutron sources as well as the most suitable system for achieving transmutation

Fast accelerator driven subcritical system for energy production: nuclear fuel evolution

International Nuclear Information System (INIS)

Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria A.F.; Costa, Antonella L.

2011-01-01

Accelerators Driven Systems (ADS) are an innovative type of nuclear system, which is useful for long-lived fission product transmutation and fuel regeneration. The ADS consist of a coupling of a sub-critical nuclear core reactor and a proton beam produced by a particle accelerator. These particles are injected into a target for the neutrons production by spallation reactions. The neutrons are then used to maintain the fission chain in the sub-critical core. The aim of this study is to investigate the nuclear fuel evolution of a lead cooled accelerator driven system used for energy production. The fuel studied is a mixture based upon "2"3"2Th and "2"3"3U. Since thorium is an abundant fertile material, there is hope for the thorium-cycle fuels for an accelerator driven sub-critical system. The target is a lead spallation target and the core is filled with a hexagonal lattice. High energy neutrons are used to reduce the negative reactivity caused by the presence of protoactinium, since this effect is most pronounced in the thermal range of the neutron spectrum. For that reason, such material is not added moderator to the system. In this work is used the Monte Carlo code MCNPX 2.6.0, that presents the the depletion/ burnup capability. The k_e_f_f evolution, the neutron energy spectrum in the core and the nuclear fuel evolution using ADS source (SDEF) and kcode-mode are evaluated during the burnup. (author)

Target/blanket conceptual design for the Los Alamos ATW concept

International Nuclear Information System (INIS)

Ames, K.; Cappiello, M.; Ireland, J.; Sapir, J.; Farnum, G.

1992-01-01

The Los Alamos Accelerator Transmutation of Waste (ATW) concept has many potential applications that include defense waste transmutation, defense material production (i.e., tritium and 238 Pu), and the transmutation of hazardous nuclear wastes from commercial nuclear reactors (fission products and actinides). A more advanced long-term Los Alamos effort is investigating the potential of an accelerator- driven system to produce fission energy with a minimal nuclear waste stream. All applications employ a high-energy (800- to 1600-MeV), high-current (25--250 mA) proton linear accelerator as the driver. In this report, we discuss only the target/blanket conceptual design for the commercial nuclear waste application. A conceptual design for the target/blanket of the Los Alamos ATW concept has been presented. The neutronics, mechanical design, and heat transfer have been investigated in some detail for the base-case design. Much more work needs to be done, but at this point it appears that the design is feasible and will approach the design goal of supporting two commercial power reactors with each target/blanket module

Neutron Detector Signal Processing to Calculate the Effective Neutron Multiplication Factor of Subcritical Assemblies

International Nuclear Information System (INIS)

Talamo, Alberto; Gohar, Yousry

2016-01-01

This report describes different methodologies to calculate the effective neutron multiplication factor of subcritical assemblies by processing the neutron detector signals using MATLAB scripts. The subcritical assembly can be driven either by a spontaneous fission neutron source (e.g. californium) or by a neutron source generated from the interactions of accelerated particles with target materials. In the latter case, when the particle accelerator operates in a pulsed mode, the signals are typically stored into two files. One file contains the time when neutron reactions occur and the other contains the times when the neutron pulses start. In both files, the time is given by an integer representing the number of time bins since the start of the counting. These signal files are used to construct the neutron count distribution from a single neutron pulse. The built-in functions of MATLAB are used to calculate the effective neutron multiplication factor through the application of the prompt decay fitting or the area method to the neutron count distribution. If the subcritical assembly is driven by a spontaneous fission neutron source, then the effective multiplication factor can be evaluated either using the prompt neutron decay constant obtained from Rossi or Feynman distributions or the Modified Source Multiplication (MSM) method.

Neutron Detector Signal Processing to Calculate the Effective Neutron Multiplication Factor of Subcritical Assemblies

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gohar, Yousry [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

2016-06-01

This report describes different methodologies to calculate the effective neutron multiplication factor of subcritical assemblies by processing the neutron detector signals using MATLAB scripts. The subcritical assembly can be driven either by a spontaneous fission neutron source (e.g. californium) or by a neutron source generated from the interactions of accelerated particles with target materials. In the latter case, when the particle accelerator operates in a pulsed mode, the signals are typically stored into two files. One file contains the time when neutron reactions occur and the other contains the times when the neutron pulses start. In both files, the time is given by an integer representing the number of time bins since the start of the counting. These signal files are used to construct the neutron count distribution from a single neutron pulse. The built-in functions of MATLAB are used to calculate the effective neutron multiplication factor through the application of the prompt decay fitting or the area method to the neutron count distribution. If the subcritical assembly is driven by a spontaneous fission neutron source, then the effective multiplication factor can be evaluated either using the prompt neutron decay constant obtained from Rossi or Feynman distributions or the Modified Source Multiplication (MSM) method.

Project 'Installation of a stand at the horizontal channel of the MARIA Research Reactor, Otwock-Swierk, Poland, for the research of transmutation of minor actinides and fission products'

International Nuclear Information System (INIS)

Szuta, M.

2006-01-01

international reputation in the field of energy production and transmutation and reactor physics computations. In order to analyse this topic experimentally, we propose to replace the expensive spallation source requiring accelerator by the neutron source obtained by converting the thermal neutron flux from the horizontal channel of the research reactor MARIA into fast neutron source. Taking into account the large amount of thermal neutrons in the horizontal channel, it is possible to use a fission converter i.e. an arrangement containing 235 U placed in the axis of the horizontal channel mouth. Thermal neutrons cause the fission reactions producing fast neutrons, which will be used instead of the neutrons from the spallation source. Five fuel rods of EK-10 type fuel placed vertically respect the beam of the thermal neutron flux of the used horizontal channel constitutes the converter. From the preliminary calculations it follows that the fast neutron source is approximately equal to 2 x 10 10 neutrons/s. A natural metallic uranium blanket with a moderator island will surround the fast neutron source what will enable us to perform the transmutation investigation of minor actinides (MA) and the long lived fission products (LLFP) in a wide range neutron energy spectrum. In such a system the mass of the 235 U will be deeply sub-critical

Preliminary analyses of neutronics schemes for three kinds waste transmutation blankets of fusion-fission hybrid

International Nuclear Information System (INIS)

Zhang Mingchun; Feng Kaiming; Li Zaixin; Zhao Fengchao

2012-01-01

The neutronics schemes of the helium-cooled waste transmutation blanket, sodium-cooled waste transmutation blanket and FLiBe-cooled waste transmutation blanket were preliminarily calculated and analysed by using the spheroidal tokamak (ST) plasma configuration. The neutronics properties of these blankets' were compared and analyzed. The results show that for the transmutation of "2"3"7Np, FLiBe-cooled waste transmutation blanket has the most superior transmutation performance. The calculation results of the helium-cooled waste transmutation blanket show that this transmutation blanket can run on a steady effective multiplication factor (k_e_f_f), steady power (P), and steady tritium production rate (TBR) state for a long operating time (9.62 years) by change "2"3"7Np's initial loading rate of the minor actinides (MA). (authors)

Comparison between two gas-cooled TRU burner subcritical reactors: fusion-fission and ADS

International Nuclear Information System (INIS)

Carluccio, T.; Rossi, P.C.R.; Angelo, G.; Maiorino, J.R.

2011-01-01

This work shows a preliminary comparative study between two gas cooled subcritical fast reactor as dedicated transuranics (TRU) transmuters: using a spallation neutron source or a D-T fusion neutron source based on ITER. The two concepts are compared in terms of a minor actinides burning performance. Further investigations are required to choose the best partition and transmutation strategy. Mainly due to geometric factors, the ADS shows better neutron multiplication. Other designs, like SABR and lead cooled ADS may show better performances than a Gas Coolead Subcritical Fast Reactors and should be investigated. We noticed that both designs can be utilized to transmutation. Besides the diverse source neutron spectra, we may notice that the geometric design and cycle parameters play a more important role. (author)

Application of the Modified Source Multiplication (MSM) Technique to Subcritical Reactivity Worth Measurements in Thermal and Fast Reactor Systems

International Nuclear Information System (INIS)

Blaise, P.; Fougeras, Ph.; Mellier, F.

2011-01-01

The Amplified Source Multiplication (ASM) method and its improved Modified Source Multiplication (MSM) method have been widely used in the CEA's EOLE and MASURCA critical facilities over the past decades for the determination of reactivity worths by using fission chambers in subcritical configurations. The ASM methodology uses relatively simple relationships between count rates of efficient miniature fission chambers located in slightly subcritical reference and perturbed configurations. While this method works quite well for small reactivity variations, the raw results need to be corrected to take into account the flux perturbation at the fission chamber location. This is performed by applying to the measurement a correction factor called MSM. This paper describes in detail both methodologies, with their associated uncertainties. Applications on absorber cluster worth in the MISTRAL-4 full MOX mock-up core and the last core loaded in MASURCA show the importance of the MSM correction on raw ASM data. (authors)

Advanced nuclear fuel production by using fission-fusion hybrid reactor

International Nuclear Information System (INIS)

Al-Kusayer, T.A.; Sahin, S.; Abdulraoof, M.

1993-01-01

Efforts are made at the College of Engineering, King Saud University, Riyadh to lay out the main structure of a prototype experimental fusion and fusion-fission (hybrid) reactor blanket in cylindrical geometry. The geometry is consistent with most of the current fusion and hybrid reactor design concepts in respect of the neutronic considerations. Characteristics of the fusion chamber, fusion neutrons and the blanket are provided. The studies have further shown that 1 GWe fission-fusion reactor can produce up to 957 kg/year which is enough to fuel five light water reactors of comparable power. Fuel production can be increased further. 29 refs

Aqueous self-cooled blanket concepts for fusion reactors

International Nuclear Information System (INIS)

Varsamis, G.; Embrechts, M.J.; Steiner, D.; Deutsch, L.; Gierszewski, P.

1987-01-01

A novel aqueous self-cooled blanket (ASCB) concept has been proposed. The water coolant also serves as the tritium breeding medium by dissolving small amounts of lithium compound in the water. The tritium recovery requirements of the ASCB concept may be facilitated by the novel in-situ radiolytic tritium separation technique in development at Chalk River Nuclear Laboratories. In this separation process deuterium gas is bubbled through the blanket coolant. Due to radiation induced processes, the equilibrium constant favors tritium migration to the deuterium gas stream. It is expected that the inherent simplicity of this design will result in a highly reliable, safe and economically attractive breeding blanket for fusion reactors. The available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate the proposed blanket concept. Tests for tritium separation and corrosion compatibility show encouraging results for the feasibility of this concept

Conceptual study on high performance blanket in a spherical tokamak fusion-driven transmuter

International Nuclear Information System (INIS)

Chen Yixue; Wu Yican

2000-01-01

A preliminary conceptual design on high performance dual-cooled blanket of fusion-driven transmuter is presented based on neutronic calculation. The dual-cooled system has some attractive advantages when utilized in transmutation of HLW (High Level Wastes). The calculation results show that this kind of blanket could safely transmute about 6 ton minor actinides (produced by 170 GW(e) Year PWRs approximately) and 0.4 ton fission products per year, and output 12 GW thermal power. In addition, the variation of power and critical factor of this blanket is relatively little during its 1-year operation period. This blanket is also tritium self-sustainable

Estimation of subcriticality by neutron source multiplication method

International Nuclear Information System (INIS)

Sakurai, Kiyoshi; Suzaki, Takenori; Arakawa, Takuya; Naito, Yoshitaka

1995-03-01

Subcritical cores were constructed in a core tank of the TCA by arraying 2.6% enriched UO 2 fuel rods into nxn square lattices of 1.956 cm pitch. Vertical distributions of the neutron count rates for the fifteen subcritical cores (n=17, 16, 14, 11, 8) with different water levels were measured at 5 cm interval with 235 U micro-fission counters at the in-core and out-core positions arranging a 252 C f neutron source at near core center. The continuous energy Monte Carlo code MCNP-4A was used for the calculation of neutron multiplication factors and neutron count rates. In this study, important conclusions are as follows: (1) Differences of neutron multiplication factors resulted from exponential experiment and MCNP-4A are below 1% in most cases. (2) Standard deviations of neutron count rates calculated from MCNP-4A with 500000 histories are 5-8%. The calculated neutron count rates are consistent with the measured one. (author)

Design of ITER neutron monitor using micro fission chambers

International Nuclear Information System (INIS)

Nishitani, Takeo; Ebisawa, Katsuyuki; Ando, Toshiro; Kasai, Satoshi; Johnson, L.C.; Walker, C.

1998-08-01

We are designing micro fission chambers, which are pencil size gas counters with fissile material inside, to be installed in the vacuum vessel as neutron flux monitors for ITER. We found that the 238 U micro fission chambers are not suitable because the detection efficiency will increase up to 50% in the ITER life time by breading 239 Pu. We propose to install 235 U micro fission chambers on the front side of the back plate in the gap between adjacent blanket modules and behind the blankets at 10 poloidal locations. One chamber will be installed in the divertor cassette just under the dome. Employing both pulse counting mode and Campbelling mode in the electronics, we can accomplish the ITER requirement of 10 7 dynamic range with 1 ms temporal resolution, and eliminate the effect of gamma-rays. We demonstrate by neutron Monte Carlo calculation with three-dimensional modeling that we avoid those detection efficiency changes by installing micro fission chambers at several poloidal locations inside the vacuum vessel. (author)

Enhanced fuel production in thorium fusion hybrid blankets utilizing uranium multipliers

International Nuclear Information System (INIS)

Pitulski, R.H.; Chapin, D.L.; Klevans, E.

1979-01-01

The multiplication of 14 MeV D-T fusion neutrons via (n,2n), (n,3n), and fission reactions by 238 U is well known and established. This study consistently evaluates the effectiveness of a depleted (tails) UO 2 multiplier on increasing the production of 233 U and tritium in a thorium/lithium fusion--fission hybrid blanket. Nuclear performance is evaluated as a function of exposure and zone thickness

Workshop summaries for the third US/USSR symposium on fusion-fission reactors

International Nuclear Information System (INIS)

Jassby, D.L.

1979-07-01

Workshop summaries on topics related to the near-term development requirements for fusion-fission (hybrid) reactors are presented. The summary topics are as follows: (1) external factors, (2) plasma engineering, (3) ICF hybrid reactors, (4) blanket design, (5) materials and tritium, and (6) blanket engineering development requirements

The Chain-Length Distribution in Subcritical Systems

International Nuclear Information System (INIS)

Nolen, Steven Douglas

2000-01-01

The individual fission chains that appear in any neutron multiplying system provide a means, via neutron noise analysis, to unlock a wealth of information regarding the nature of the system. This work begins by determining the probability density distributions for fission chain lengths in zero-dimensional systems over a range of prompt neutron multiplication constant (K) values. This section is followed by showing how the integral representation of the chain-length distribution can be used to obtain an estimate of the system's subcritical prompt multiplication (MP). The lifetime of the chains is then used to provide a basis for determining whether a neutron noise analysis will be successful in assessing the neutron multiplication constant, k, of the system in the presence of a strong intrinsic source. A Monte Carlo transport code, MC++, is used to model the evolution of the individual fission chains and to determine how they are influenced by spatial effects. The dissertation concludes by demonstrating how experimental validation of certain global system parameters by neutron noise analysis may be precluded in situations in which the system K is relatively low and in which realistic detector efficiencies are simulated

The Chain-Length Distribution in Subcritical Systems

Energy Technology Data Exchange (ETDEWEB)

Nolen, Steven Douglas [Texas A & M Univ., College Station, TX (United States)

2000-06-01

The individual fission chains that appear in any neutron multiplying system provide a means, via neutron noise analysis, to unlock a wealth of information regarding the nature of the system. This work begins by determining the probability density distributions for fission chain lengths in zero-dimensional systems over a range of prompt neutron multiplication constant (K) values. This section is followed by showing how the integral representation of the chain-length distribution can be used to obtain an estimate of the system's subcritical prompt multiplication (MP). The lifetime of the chains is then used to provide a basis for determining whether a neutron noise analysis will be successful in assessing the neutron multiplication constant, k, of the system in the presence of a strong intrinsic source. A Monte Carlo transport code, MC++, is used to model the evolution of the individual fission chains and to determine how they are influenced by spatial effects. The dissertation concludes by demonstrating how experimental validation of certain global system parameters by neutron noise analysis may be precluded in situations in which the system K is relatively low and in which realistic detector efficiencies are simulated.

Direct LiT Electrolysis in a Metallic Fusion Blanket

Energy Technology Data Exchange (ETDEWEB)

Olson, Luke [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-09-30

A process that simplifies the extraction of tritium from molten lithium-based breeding blankets was developed. The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fusion/fission reactors is critical in order to maintain low concentrations. This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Extraction is complicated due to required low tritium concentration limits and because of the high affinity of tritium for the blanket. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering hydrogen and deuterium through an electrolysis step at high temperatures.

Direct LiT Electrolysis in a Metallic Fusion Blanket

International Nuclear Information System (INIS)

Olson, Luke

2016-01-01

A process that simplifies the extraction of tritium from molten lithium-based breeding blankets was developed. The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fusion/fission reactors is critical in order to maintain low concentrations. This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Extraction is complicated due to required low tritium concentration limits and because of the high affinity of tritium for the blanket. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering hydrogen and deuterium through an electrolysis step at high temperatures.

Development of reactivity feedback effect measurement techniques under sub-critical condition in fast reactors

International Nuclear Information System (INIS)

Kitano, A.; Nishi, H.; Suzuki, T.; Okajima, S.; Kanemoto, S.

2012-01-01

The first-of-a-kind reactor has been licensed by a safety examination of the plant design based on the measured data in precedent mock-up experiments. The validity of the safety design can be confirmed without a mock-up experiment, if the reactor feed-back characteristics can be measured before operation, with the constructed reactor itself. The 'Synthesis Method', a systematic and sophisticated method of sub-criticality measurement, is proposed in this work to ensure the safety margin before operation. The 'Synthesis Method' is based on the modified source multiplication method (MSM) combined with the noise analysis method to measure the reference sub-criticality level for MSM. A numerical simulation for the control-rod reactivity worth and the isothermal feed-back reactivity was conducted for typical fast reactors of 100 MWe-size, 300 MWe-size, 750 MWe-size, and 1500 MWe-size to investigate the applicability of Synthesis Method. The number of neutron detectors and their positions necessary for the measurement were investigated for both methods of MSM and the noise analysis by a series of parametric survey calculations. As a result, it was suggested that a neutron detector located above the core center and three or more neutron detectors located above the radial blanket region enable the measurement of sub-criticality within 10% uncertainty from -$0.5 to -$2 and within 15% uncertainty for the deeper sub-criticality. (authors)

Accelerator-driven thermal fission systems may provide energy supply advantages

International Nuclear Information System (INIS)

Linford, R.K.

1992-01-01

This presentation discusses the energy supply advantages of using accelerator-driven thermal fission systems. Energy supply issues as related to cost, fuel supply stability, environmental impact, and safety are reviewed. It is concluded that the Los Alamos Accelerator Transmutation of Waste (ATW) concept, discussed here, has the following advantages: improved safety in the form of low inventory and subcriticality; reduced high-level radioactive waste management timescales for both fission products and actinides; and a very long-term fuel supply requiring no enrichment

Implementation and training methodology of subcritical reactors neutronic calculations triggered by external neutron source and applications

International Nuclear Information System (INIS)

Carluccio, Thiago

2011-01-01

This works had as goal to investigate calculational methodologies on subcritical source driven reactor, such as Accelerator Driven Subcritical Reactor (ADSR) and Fusion Driven Subcritical Reactor (FDSR). Intense R and D has been done about these subcritical concepts, mainly due to Minor Actinides (MA) and Long Lived Fission Products (LLFP) transmutation possibilities. In this work, particular emphasis has been given to: (1) complement and improve calculation methodology with neutronic transmutation and decay capabilities and implement it computationally, (2) utilization of this methodology in the Coordinated Research Project (CRP) of the International Atomic Energy Agency Analytical and Experimental Benchmark Analysis of ADS and in the Collaborative Work on Use of Low Enriched Uranium in ADS, especially in the reproduction of the experimental results of the Yalina Booster subcritical assembly and study of a subcritical core of IPEN / MB-01 reactor, (3) to compare different nuclear data libraries calculation of integral parameters, such as k eff and k src , and differential distributions, such as spectrum and flux, and nuclides inventories and (4) apply the develop methodology in a study that may help future choices about dedicated transmutation system. The following tools have been used in this work: MCNP (Monte Carlo N particle transport code), MCB (enhanced version of MCNP that allows burnup calculation) and NJOY to process nuclear data from evaluated nuclear data files. (author)

Neutron measurements in the core and blankets of the reactor Rapsodie

International Nuclear Information System (INIS)

Gourdon, J.; Edeline, J.C.

1968-01-01

Beside a brief general discussion, the report contains all the core and blanket neutronic measurements. It covers successively the methods, the measurements themselves and the results. The later concern: spectral indexes, axial and radial fission rates, activation foil measurements and neutronic power determination. (authors) [fr

{open_quotes}Transmutation efficiency calculation in the blanket on melted salts with central neptunium target{close_quotes}

Energy Technology Data Exchange (ETDEWEB)

Kolesov, V.F.; Shtarev, S.K.; Khoruzhiy, V.K. [Russia Federal Nuclear Center, Arzamas (Russian Federation)] [and others

1995-10-01

In the limits of ABC project version of two-sectional reactor system in the form of combination of subcritical blanket on melted salts and multiplying target from threshold fissile material {sup 237}Np is considered. This research is the development of the VNIIEF`s earlier work`s (Russia) investigating of usage possibilities in ABC project the conception of multisectional blankets with single-sided neutron coupling between sections. With the help of Monte-Carlo program the calculations results of system mentioned are given. The possibility of accelerator`s considerable power reduction at the account of thorium target substitution with neptunium-237 multiplying target is shown.

Feasibility study of a fission-suppressed tandem-mirror hybrid reactor

International Nuclear Information System (INIS)

Lee, J.D.; Moir, R.W.; Barr, W.L.

1982-04-01

Results of a conceptual design study of a U-233 producing fusion breeder consisting of a tandem mirror fusion device and two types of fission-suppressed blankets are presented. The majority of the study was devoted to the conceptual design and evaluation of the two blankets. However, studies in the areas of fusion engineering, reactor safety, fuel reprocessing, other fuel cycle issues, economics, and deployment were also performed

Present status of the EPFL (Swiss) fusion-fission experiment 'LOTUS'

International Nuclear Information System (INIS)

Haldy, P.A.; Frueh, R.; Ligou, J.; Schneeberger, J.P.; Kumar, A.

1984-01-01

The present status of the LOTUS project - a fusion-fission hybrid research facility under construction at the Ecole Polytechnique Federale de Lausanne (EPFL) Switzerland - is presented. Emphasis is places on the description of the facility and on the design studies of an initial blanket of the ''fission-suppressed'' type. The LOTUS facility consists of a parallelepiped-shaped blanket, occupying roughly a volume of 1 m 3 , driven by a sealed 14 MeV (D,T) neutron generator with a rated source strength of 5x10 12 n/s. The experiment is housed in a massive concrete shielding of 220 cm thick walls, which leaves an experimental test chamber of 360 cm by 240 cm lateral dimensions and a height of 300 cm. (orig.) [de

Technical issues for beryllium use in fusion blanket applications

International Nuclear Information System (INIS)

McCarville, T.J.; Berwald, D.H.; Wolfer, W.; Fulton, F.J.; Lee, J.D.; Maninger, R.C.; Moir, R.W.; Beeston, J.M.; Miller, L.G.

1985-01-01

Beryllium is an excellent non-fissioning neutron multiplier for fusion breeder and fusion electric blanket applications. This report is a compilation of information related to the use of beryllium with primary emphasis on the fusion breeder application. Beryllium resources, production, fabrication, properties, radiation damage and activation are discussed. A new theoretical model for beryllium swelling is presented

Design of In-vessel neutron monitor using micro fission chambers for ITER

International Nuclear Information System (INIS)

Nishitani, Takeo; Kasai, Satoshi

2001-10-01

A neutron monitor using micro fission chambers to be installed inside the vacuum vessel has been designed for compact ITER (ITER-FEAT). We investigated the responses of the micro fission chambers to find the suitable position of micro fission chambers by a neutron Monte Carlo calculation using MCNP version 4b code. It was found that the averaged output of the micro fission chambers behind blankets at upper outboard and lower outboard is insensitive to the changes in the plasma position and the neutron source profile. A set of 235 U micro fission chamber and ''blank'' detector which is a fissile material free detector to identify noise issues such as from γ-rays are installed behind blankets. Employing both pulse counting mode and Campbelling mode in the electronics, the ITER requirement of 10 7 dynamic range with 1 ms temporal resolution can be accomplished. The in-situ calibration has been simulated by MCNP calculation, where a point source of 14 MeV neutrons is moving on the plasma axis. It was found that the direct calibration is possible by using a neutron generator with an intensity of 10 11 n/s. The micro fission chamber system can meet the required 10% accuracy for a fusion power monitor. (author)

Role of fission-reactor-testing capabilities in the development of fusion technology

International Nuclear Information System (INIS)

Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

1981-01-01

Testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for testing when fusion reactors become available. Fission testing is capable of filling many gaps in fusion reactor design information, and thus should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, we investigated radiation damage to magnet insulators. This work is now continuing with the use of an improved test capsule. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, we explored a conceptual design for a fission-based Integrated Test Facility (ITF), which can accommodate entire First Wall/Blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

233U breeding in accelerator-driven sub-critical fast reactor

International Nuclear Information System (INIS)

Yang Yongwei; An Yu

1999-01-01

Accelerator-driven Sub-critical Fast Reactor (ADFR) is chosen as fissile-material-breeding reactor. (U-Pu)O x is chosen as fuel in the core and ThO 2 as fertile material in the blanket zone to breed 233 U. Molten lead is chosen as coolant because of its better neutronic and chemical characteristics over sodium. The program system used for neutronics study consists of: LAHET, for the simulation of the interaction between the proton with medium energy and the nuclei of the target; MCNP4A, for the simulation of neutron transport with energy below 20 MeV in the sub-critical reactor; CONNECT1, for the processing of some tallies provided by the output of MCNP4A in order to prepare micro-cross sections for elements used for burnup calculation; ORIGEN2, used for multi-region burnup calculation; CONNECT2, for the processing of atom densities of some elements provided in the output of ORIGEN2 in order to prepare input to LAHET calculation for next time step. The calculated results show that the proposed case is feasible for breeding fissile material considering the criticality safety, power density, burnup, etc

Application of the Modified Source Multiplication (MSM) technique to subcritical reactivity worth measurements in thermal and fast reactor systems

International Nuclear Information System (INIS)

Blaise, P.; Fougeras, P.; Mellier, F.

2009-01-01

The Amplified Source Multiplication (ASM) method and its improved Modified Source Multiplication (MSM) method have been widely used in the CEA's EOLE and MASURCA critical facilities over the past decades for the determination of reactivity worths by using fission chambers in subcritical configurations. They have been successfully applied to absorber (single or clusters) worth measurement in both thermal and fast spectra, or for (sodium or water) void reactivity worths. The ASM methodology, which is the basic technique to estimate a reactivity worth, uses relatively simple relationships between count rates of efficient miniature fission chambers located in slightly subcritical reference and perturbed configurations. If this method works quite well for small reactivity variation (a few effective delayed neutron fraction), its raw results needs to be corrected to take into account the flux perturbation in the fission chamber. This is performed by applying to the measurement a correction factor called MSM. Its characteristics is to take into account the local space and energy variation of the spectrum in the fission chamber, through standard perturbation theory applied to neutron transport calculation in the perturbed configuration. The proposed paper describes in details both methodologies, with their associated uncertainties. Applications on absorber cluster worth in the MISTRAL-4 full MOX mock-up core and the last core loaded in MASURCA show the importance of the MSM correction on raw data. (authors)

Burnup calculations for KIPT accelerator driven subcritical facility using Monte Carlo computer codes-MCB and MCNPX

International Nuclear Information System (INIS)

Gohar, Y.; Zhong, Z.; Talamo, A.

2009-01-01

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron accelerator. The neutron source of the subcritical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and electron energy in the range of 100 to 200 MeV. The main functions of the subcritical assembly are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron physics experiments and material structure analyses are planned using this facility. With the 100 KW electron beam power, the total thermal power of the facility is ∼375 kW including the fission power of ∼260 kW. The burnup of the fissile materials and the buildup of fission products reduce continuously the reactivity during the operation, which reduces the neutron flux level and consequently the facility performance. To preserve the neutron flux level during the operation, fuel assemblies should be added after long operating periods to compensate for the lost reactivity. This process requires accurate prediction of the fuel burnup, the decay behavior of the fission produces, and the introduced reactivity from adding fresh fuel assemblies. The recent developments of the Monte Carlo computer codes, the high speed capability of the computer processors, and the parallel computation techniques made it possible to perform three-dimensional detailed burnup simulations. A full detailed three-dimensional geometrical model is used for the burnup simulations with continuous energy nuclear data libraries for the transport calculations and 63-multigroup or one group cross sections libraries for the depletion calculations. Monte Carlo Computer code MCNPX and MCB are utilized for this study. MCNPX transports the electrons and the

Hybrid nuclear cycles for nuclear fission sustainability

International Nuclear Information System (INIS)

Piera, M.; Martinez-Val, M. M.

2007-01-01

Nuclear fission can play and must play an important role in paving the road to Energy Sustainability. Nuclear Fission does not produce CO 2 emissions, and it is already exploited at commercial level with the current NPP (Nuclear Power Plants). Most of them are based on LWR reactors, which have a very good safety record. It must be noted, however, that all LWR (including the advanced or evolutionary ones) have some drawbacks, particularly their very poor efficiency in exploiting the natural resources of nuclear fuels. In this paper, an analysis is presented on how to maximize the energy actually generated from the potential contents of fission natural resources. The role of fertile-to-fissile breeding is highlighted, as well as the need of attaining a very high safety performance in the reactors and other installations of the fuel cycle. The proposal presented in this paper is to use advanced and evolutionary LWR as energy producing reactors, and to use subcritical fast assemblies as breeders. The main result would be to increase by two orders of magnitude the percentage of energy effectively exploited from fission natural resources, while keeping a very high level of safety standards in the full fuel cycle. Breeders would not be intended for energy production, so that safety standards could rely on very low values of the thermal magnitudes, so allowing for very large safety margins for emergency cooling. Similarly, subcriticality would offer a very large margin for not to reach prompt criticality in any event. The main drawback of this proposal is that a sizeable fraction of the energy generated in the cycle (about 1/3, maybe a little more) would not be useful for the thermodynamic cycle to produce electricity. Besides that, a fraction of the generated electricity, between 5 and 10 %, would have to be recirculated to feed the accelerator activating the neutron source. Even so, the overall result would be very positive, because more than 50 % of the natural

Measurements of Pu239:U235 fission ratio using foils at temperatures up to 400 deg. C

Energy Technology Data Exchange (ETDEWEB)

Carter, D H; Puckett, B J; Richards, A E [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1964-05-15

The paper describes the use of activation foils for the measurement of Pu239:U235 fission ratios in subcritical lattices at temperatures up to 390 deg C. Counting techniques and the method of analysis of the results are described in detail and the results are compared with fission chamber measurements. (author) 4 refs., 6 figs., 7 tabs.

Some applications of fission-based testing capabilities in the development of fusion technology

International Nuclear Information System (INIS)

Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

1981-10-01

The testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for fusion testing when they do become available. Fission testing is capable of filling many gaps in fusion reactor design information, and should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, work was performed on the irradiation of magnet insulators. This work is continuing with an improved test environment. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, a conceptual design is presented for a fission-based Integrated Test Facility (ITF), which can accommodate entire wall/blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

Economic regimes for fission--fusion energy systems

International Nuclear Information System (INIS)

Deonigi, D.E.

1974-01-01

The objectives of this hybrid fusion-fission economic regimes study are to: (1) define the target costs to be met, (2) define the optimum fissile/electrical production ratio for hybrid blankets, (3) discover synergistic configurations, and (4) define the windows of economic hybrid design having desirable cost/benefit ratios. (U.S.)

Choice of economical optimum blanket of hybrid reactors

Energy Technology Data Exchange (ETDEWEB)

Blinkin, V L; Novikov, V M

1981-01-01

The economical effectiveness of symbiotic power systems depends on the choice of the correlation between energy production and fissile fuel production in blankets of controlled thermonuclear fusion reactor (CTR), what is investigated here. It is shown that the optimum value of this correlation essentially depends on the ratio between the specific costs for energy production in hybrid thermonuclear reactors and that in fission reactors as part of the symbiotic system.

Neutron noise measurements at the Delphi subcritical assembly

International Nuclear Information System (INIS)

Szieberth, M.; Klujber, G.; Kloosterman, J. L.; De Haas, D.

2012-01-01

The paper presents the results and evaluations of a comprehensive set of neutron noise measurements on the Delphi subcritical assembly of the Delft Univ. of Technology. The measurements investigated the effect of different source distributions (inherent spontaneous fission and 252 Cf) and the position of the detectors applied (both radially and vertically). The evaluation of the measured data has been performed by the variance-to-mean ratio (VTMR, Feynman-α), the autocorrelation (ACF, Rossi-α) and the cross-correlation (CCF) methods. The values obtained for the prompt decay constant show a strong bias, which depends both on the detector position and on the source distribution. This is due to the presence of higher modes in the system. It has been observed that the α value fitted is higher when the detector is close to the boundary of the core or to the 252 Cf point-source. The higher alpha-modes have also been observed by fitting functions describing two alpha-modes. The successful set of measurement also provides a good basis for further theoretical investigations including the Monte Carlo simulation of the noise measurements and the calculation of the alpha-modes in the Delphi subcritical assembly. (authors)

Subcriticality determination of nuclear reactor

International Nuclear Information System (INIS)

Borisenko, V.I.; Goranchuk, V.V.; Sidoruk, N.M.; Volokh, A.F.

2014-01-01

In this article the subcriticality determination of nuclear reactor is considered. Emphasized that, despite the requirements of regulatory documents on the subcriticality determination of WWER from the beginning of their operation, so far, this problem has not been solved. The results of subcriticality determination of Rossi-α method of the WWER-M is presented. The possibility of subcriticality determination of WWER is considered. The possibility of subcriticality determination of Rossi-α method with time resolution is of about 100 microseconds is also considered. The possible reasons for the error in subcriticality determination of the reactor are indicated

Blankets for fusion reactors : materials and neutronics

International Nuclear Information System (INIS)

Carvalho, S.H. de.

1980-03-01

The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author) [pt

Fusion-fission hybrid studies in the United States

International Nuclear Information System (INIS)

Moir, R.W.; Lee, J.D.; Berwald, D.H.; Cheng, E.T.; Delene, J.G.; Jassby, D.L.

1986-01-01

Systems and conceptual design studies have been carried out on the following three hybrid types: (1) The fission-suppressed hybrid, which maximizes fissile material produced (Pu or 233 U) per unit of total nuclear power by suppressing the fission process and multiplying neutrons by (n,2n) reactions in materials like beryllium. (2) The fast-fission hybrid, which maximizes fissile material produced per unit of fusion power by maximizing fission of 238 U (Pu is produced) in which twice the fissile atoms per unit of fusion power (but only a third per unit of nuclear power) are made. (3) The power hybrid, which amplifies power in the blanket for power production but does not produce fuel to sell. All three types must sell electrical power to be economical

Fast Breeder Blanket Facility FBBF. Annual report, January 1, 1981-December 31, 1981

International Nuclear Information System (INIS)

Clikeman, F.M.

1982-07-01

This annual report contains a summmary of fission rate, spectra, and gamma-ray heating rate measurements made in the first blanket of the Purdue Fast Breeder Blanket Facility. The first blanket consisted of aluminum clad, natural UO 2 fuel rods with a secondary cladding of stainless steel or aluminum. The blanket was arranged in two concentric regions around the neutron source and converter regions. A neutron diffusion code, 2DB, and a Monte Carlo code, VIM, both using homogeneous cross section groups have been used to calculate the reaction rates. Calculated to experimental values for a number of important reactions are presented. A modified method of applying Bondarenko self-shielding factors to correct for the self shielding of resonance energy neutrons in aluminum, stainless steel and UO 2 has improved the agreement between the calculations and experiment, but does not account for all of the differences

Optimization of the fission--fusion hybrid concept

International Nuclear Information System (INIS)

Saltmarsh, M.J.; Grimes, W.R.; Santoro, R.T.

1979-04-01

One of the potentially attractive applications of controlled thermonuclear fusion is the fission--fusion hybrid concept. In this report we examine the possible role of the hybrid as a fissile fuel producer. We parameterize the advantages of the concept in terms of the performance of the fusion device and the breeding blanket and discuss some of the more troublesome features of existing design studies. The analysis suggests that hybrids based on deuterium--tritium (D--T) fusion devices are unlikely to be economically attractive and that they present formidable blanket technology problems. We suggest an alternative approach based on a semicatalyzed deuterium--deuterium (D--D) fusion reactor and a molten salt blanket. This concept is shown to emphasize the desirable features of the hybrid, to have considerably greater economic potential, and to mitigate many of the disadvantages of D--T-based systems

Accelerator driven systems (ADS): A principal neutronics and transmutation potential

International Nuclear Information System (INIS)

Slessarev, I.

1997-01-01

An accelerator-based system using a beam of high energy protons to produce supplementary neutrons as a result of spallation processes in a target is investigated. The spallation neutrons are successively used to feed a subcritical blanket where they create a neutron surplus available for incineration of those long-lived toxic nuclei which require neutrons (long-lived fission products and minor actinides), and enhance the deterministic safety features for reactivity-type of accidents

MIT LMFBR blanket physics project progress report No. 7, July 1, 1975--September 30, 1976

International Nuclear Information System (INIS)

Driscoll, M.J.

1976-01-01

Work during the period was devoted primarily to a range of analytical/numerical investigations, including evaluation of means to improve external blanket designs, beneficial attributes of the use of internal blankets, improved methods for the calculation of heterogeneous self-shielding and parametric studies of calculated spectral indices. Experimental work included measurements of the ratio of U-238 captures to U-235 fissions in a standard blanket mockup, and completion of development work on the radiophotoluminescent readout of LiF thermoluminescent detectors. The most significant findings were that there is very little prospect for substantial improvement in the breeding performance of external blankets, but internal blankets continue to show promise, particularly if they are used in such a way as to increase the volume fraction of fuel inside the core envelope. An improved equivalence theorem was developed which may allow use of fast reactor methods to calculate heterogeneously self-shielded cross sections in both fast and thermal reactors

Neutronic calculations for a subcritical system with external source

International Nuclear Information System (INIS)

Cintas, A; Lopasso, E.M; Marquez Damian, J. I

2006-01-01

We present a neutronic study on an A D S, systems capable of transmute minor actinides and fission products in order to reduce their radiotoxicity and mean-life.We compare neutronic parameters obtained with Scale/Tort and M C N P modelling a sub-critical system with source from a N E A Benchmark.Due to lack of nuclear data at the temperature of the system, we perform calculations at available temperature of libraries (300 K); to compensate the reactivity insertion due to the temperature change we reduce the size of the fuel zone in order to get a sub-critical system that allow u s to evaluate neutronic parameters of the system with source.We have found that the numerical results (neutron spectrum, neutron flux distributions and other neutronic parameters) are in agreement with the M C N P and with those of the benchmark participants even though the geometric models used are not exactly the same. We conclude that with the real temperature cross sections, the calculation scheme developed (Scale/Tort and M C N P) will give reliable results in A D S evaluations [es

Lead cooled heterogeneous accelerator driven molten-fluoride blanket for incineration of long-lived radioactive wastes

International Nuclear Information System (INIS)

Lopatkin, A.V.; Matyushechkin, V.M.; Tretyakov, I.T.; Blagovolin, P.P.; Kazaritsky, V.D.

1997-01-01

This paper presents a tentative design description and evaluation of the basic parameters of a lead cooled heterogeneous accelerator driven molten fluoride blanket. The proton beam of a 1 GeV accelerator strikes the blanket from below and generates spallation neutrons in the flow of lead, which serves as a target. These neutrons leave the target zone and get into a heterogeneous blanket with separated volumes of molten salts and lead. Fissile materials are dissolved in the salt. On getting into the molten salt volume the neutrons cause fission (transmutation) of the actinides, the produced heat being removed by circulation of molten lead. Two versions of the blanket design are examined. The first version: molten salt circulates in the fuel channels, while lead cools the channels flowing through the interchannel space (the salt channel design). The second version: it is lead that circulates in the channels, while molten salt takes up the interchannel space (the lead channel design). A preliminary blanket design study showed that both blanket designs possess a potential for improving performance. At present time the blanket design, mentioned above as the salt channel design, seems to be more promising. 1 ref., 2 figs., 2 tabs

Calculated characteristics of subcritical assembly with anisotropic transport of neutrons

International Nuclear Information System (INIS)

Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I.

2003-01-01

There was considered possibility of creating enough sub-critical system that multiply neutron fluence from a primary source by many orders. For assemblies with high neutron tie between parts, it is impossible. That is why there was developed a construction consisting of many units (cascades) having weak feedback with preceding cascades. The feedback attenuation was obtained placing layers of slow neutron absorber and moderators between the cascades of fission material. Anisotropy of fast neutron transport through the layers was used. The system consisted of many identical cascades aligning one by another. Each cascade consists of layers of moderator, fissile material and absorber of slow neutrons. The calculations were carried out using the code MCNP.4a with nuclear data library ENDF/B5. In this construction neutrons spread predominantly in one direction multiplying in each next fissile layer, and they attenuate considerably in the opposite direction. In a calculated construction, multiplication factor of one cascade is about 1.5 and multiplication factor of whole construction composed of n cascades is 1.5 n . Calculated keff value is 0.9 for one cascade and does not exceed 0.98 for a system containing any number of cascades. Therefore the assembly is always sub-critical and therefore it is safe in respect of criticality. There was considered using such a sub-critical assembly to create a powerful neutron fluence for neutron boron-capturing therapy. The system merits and demerits were discussed. (authors)

BRAHMMA - accelerator driven subcritical facility

International Nuclear Information System (INIS)

Roy, Tushar; Shukla, Shefali; Shukla, M.; Ray, N.K.; Kashyap, Y.S.; Patel, T.; Gadkari, S.C.

2017-01-01

Accelerator Driven Subcritical systems are being studied worldwide for their potential in burning minor actinides and reducing long term radiotoxicity of spent nuclear fuels. In order to pursue the physics studies of Accelerator Driven Subcritical systems, a thermal subcritical assembly BRAHMMA (BeOReflectedAndHDPeModeratedMultiplying Assembly) has been developed at Purnima Labs, BARC. The facility consists of two major components: Subcritical core and Accelerator (DT/ DD Purnima Neutron Generator)

Systems study of tokamak fusion--fission reactors

International Nuclear Information System (INIS)

Tenney, F.H.; Bathke, C.G.; Price, W.G. Jr.; Bohlke, W.H.; Mills, R.G.; Johnson, E.F.; Todd, A.M.M.; Buchanan, C.H.; Gralnick, S.L.

1978-11-01

This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations

Unified description of neutron-, proton- and photon-induced fission cross sections in intermediate energy region

International Nuclear Information System (INIS)

Fukahori, Tokio; Iwamoto, Osamu; Chiba, Satoshi

2003-01-01

For an accelerator-driven nuclear waste transmutation system, it is very important to estimate sub-criticality of core system for feasibility and design study of the system. The fission cross section in the intermediate energy range has an important role. A program FISCAL has been developed to calculate neutron-, proton- and photon-induced fission cross sections in the energy region from several tens of MeV to 3 GeV. FISCAL adopts the systematics considering experimental data for Ag- 243 Am. It is found that unified description of neutron-, proton- and photon-induced fission cross sections is available. (author)

Scoping studies of 233U breeding fusion fission hybrid

International Nuclear Information System (INIS)

Maniscalco, J.A.; Hansen, L.F.; Allen, W.O.

1978-05-01

Neutronic calculations have been carried out in order to design a laser fusion driven hybrid blanket which maximizes 233 U production per unit of thermal energy (greater than or equal to 1 kg/MW/sub T/-year) with acceptable fusion energy multiplication (M/sub F/ approximately 4). Two hybrid blankets, a thorium and a uranium-thorium blanket, are discussed in detail and their performance is evaluated by incorporating them into an existing hybrid design (the LLL/Bechtel design). The overall performance of the two laser fusion driven 233 U producers is discussed and estimates are given of (1) the number of equivalent thermal power fission reactors (LWR, HWR, SSCR and HTGR) that these fusion breeders can fuel, (2) their capital cost, and (3) the cost of electricity in the combined fusion breder-converter reactor scenario

Scoping studies of 233U breeding fusion fission hybrid

International Nuclear Information System (INIS)

Maniscalco, J.A.; Hansen, L.F.; Allen, W.O.

1978-01-01

Neutronic calculations have been carried out in order to design a laser fusion driven hybrid blanket which maximizes 233 U production per unit of thermal energy (greater than or equal to 1 kg/MW/sub T/-year) with acceptable fusion energy multiplication (M/sub F/ approx. 4). Two hybrid blankets, a thorium and a uranium--thorium blanket, are discussed in detail and their performance is evaluated by incorporating them into an existing hybrid design (the LLL/Bechtel design). The overall performance of the two laser fusion driven 233 U producers is discussed and estimates are given of (1) the number of equivalent thermal power fission reactors (LWR, HWR, SSCR and HTGR) that these fusion breeders can fuel, (2) their capital cost, and (3) the cost of electricity in the combined fusion breeder-converter reactor scenario

Fission-suppressed hybrid reactor: the fusion breeder

International Nuclear Information System (INIS)

Moir, R.W.; Lee, J.D.; Coops, M.S.

1982-12-01

Results of a conceptual design study of a 233 U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed

Cost optimization of ADS design: Comparative study of externally driven heterogeneous and homogeneous two-zone subcritical reactor systems

International Nuclear Information System (INIS)

Gulik, Volodymyr; Tkaczyk, Alan H.

2014-01-01

volume/volume ratio (outer zone volume to inner zone volume) may exist, with zones differing in fuel content; and this ratio corresponds to a minimal total fuel cost. Such ratio can be obtained for a two-zone subcritical reactor depending on the material composition of its zones. Based on this article a model of subcritical reactor could be developed, which would be aimed at a transmutation of minor actinides or long-lived fission products, or at specific scientific and applied objectives

Analysis of the MUSE-3 subcritical experiment

Energy Technology Data Exchange (ETDEWEB)

Aliberti, G; Rimpault, G; Jacqmin, R; Lebrat, J F; Chauvin, J P; Granget, G [CEA Cadarache, Dept. d' Etudes des Reacteurs 13 - Saint Paul lez Durance (France); Salvatores, M [CEA Saclay, Dir. de l' Energie Nucleaire, 91 - Gif sur Yvette (France)

2001-07-01

The purpose of the MUSE (MUltiplication avec Source Externe) experimental programme is to investigate the neutronic properties of fast sub-critical cores coupled with known external sources of neutrons. Measurements of the MUSE-3 experiments (third phase of the MUSE program) included reactivity, U-235 fission rates across various traverses, absolute fission rates and dynamic measurements. Special care was taken in assessing the various sources of errors and uncertainties affecting the results when modelling and analysing MUSE-3 experiments with the ERANOS neutronic code system. Measured and calculated values agree well with each other (discrepancies within the uncertainty bars) except absolute fission rates which are affected by the large uncertainties associated with the inherent source (30 %) and with the strength of the fusion source (25%). However, such uncertainties do not contribute to the uncertainty in the importance of the source, {phi}{sup *}, which is linked to the ratio of the measured reaction rate with and without the fusion source (the generator being switched off and on). The analysis yielded therefore valuable results, in particular on the relative importance of the source from one configuration to another. The uncertainty in the calculated {phi}{sup *}, is of 1-2% (JEF2 and ERALIB1) for configurations without diffuser and 6% for the configuration with a Pb diffuser. There is therefore no bias in this important ADS characteristic as calculations lie within the measured error bars. From this analysis, it can be concluded that MUSE-3-type experiments are suitable for the assessment of tools used for designing ADS. (author)

Investigation of lanthanides as neutron multipliers for hybrid and fusion reactor blankets

International Nuclear Information System (INIS)

Sahin, Sumer

1982-01-01

The neutronic performance of three lanthanides ( 149 Sm, europium, and gadolinium) as neutron multiplier for the blanket of a fusion-fission (hybrid) and a pure fusion reactor has been evaluated and compared with that of beryllium and lead. During the calculations, the fission zone is made up of UO 2 rods from the LOTUS experimental hybrid facility now under construction at the Nuclear Engineering Laboratory of the Swiss Federal Institute of Technology in Lausanne. In fusion blanket the fuel zone is replaced by pure lithium. The calculations were performed for two different boundary conditions for the left boundary: (a) reflecting, representative of a typical confinement geometry, and (b) vacuum, which represents a typical blanket experiment in plane geometry. For a vacuum left boundary, threshold reactions are reduced by a factor of about 2 while 1/v-type reactions are decreased by a factor of between 5 and 10, as a consequence of the softer spectrum produced by a reflecting left boundary. In general, the results, notably tritium breeding and energy multiplication, are comparable for the lanthanide multipliers and for beryllium and lead if the left boundary is a vacuum. The use of 149 Sm is slightly less effective than europium or gadolinium and all of the lanthanides perform better for a vacuum left boundary than for the reflecting case. The analyses presented here also illustrate the importance of potential spectral shifts that can occur as the result of experimental exigencies

Criticality safety margins for mixtures of fissionable materials

International Nuclear Information System (INIS)

Williamson, T.G.; Mincey, J.F.

1992-01-01

In the determination of criticality safety margins, approximations for combinations of fissile and fissionable isotopes are sometimes used that go by names such as the rule of fractions or equivalency relations. Use of the rule of fractions to ensure criticality safety margins was discussed in an earlier paper. The purpose of this paper is to correct errors and to clarify some of the implications. Deviations of safety margins from those calculated by the rule of fractions are still noted; however, the deviations are less severe. Caution in applying such rules is still urged. In general, these approximations are based on American National Standard ANSI/ANS-8.15, Sec. 5.2. This section allows that ratios of material masses to their limits may be summed for fissile nuclides in aqueous solutions. It also allows the addition of nonfissile nuclides if an aqueous moderator is present and addresses the effects of infinite water or equivalent reflector. Water-reflected binary combinations of aqueous solutions of fissile materials, as well as binary combinations of fissile and fissionable metals, were considered. Some combinations were shown to significantly decrease the margin of subcriticality compared to the single-unit margins. In this study, it is confirmed that some combinations of metal units in an optimum geometry may significantly decrease the margin of subcriticality. For some combinations of aqueous solutions of fissile materials, the margin of subcriticality may also be reduced by very small amounts. The conclusion of Ref. 1 that analysts should be careful in applying equivalency relations for combining materials remains valid and sound advice. The ANSI/ANS standard, which allows the use of ratios of masses to their limits, applies to aqueous, fully water-reflected, single-unit solutions. Extensions to other situations should be considered with extreme care

Nuclear fission sustainability with subcritical reactors driven by external neutron sources

International Nuclear Information System (INIS)

Lafuente, A.; Piera, M.

2011-01-01

Although nuclear breeder reactors are a promising way to enhance the potential energy currently retrievable from the Uranium reserves, they still have disadvantages because of their safety features (i.e. poor stabilizing mechanisms) and the security of their fuel cycle (diversion of Pu for non-civilian purposes). Loading natural nuclear fuels to a reactor and completely burning them without reprocessing would be ideal, however, this is not possible in critical reactors due to the limitations imposed by the maximum achievable burn-up. An alternative option to attain very high percentages of nuclear natural materials exploitation, while meeting other objectives of Nuclear Sustainability, could consist of using externally-driven subcritical reactors to reach the desired high burn-ups (of the order of 30% and more) without reprocessing. Such scheme would lead to an efficient exploitation of the available raw material, without any risk of proliferation. Exploring this type of reactor concept, this paper analyzes the different ways to accomplish this goal while identifying potential setbacks.

Conceptual design of a fission-based integrated test facility for fusion reactor components

International Nuclear Information System (INIS)

Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

1982-01-01

The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

Contained fissionly vaporized imploded fission explosive breeder reactor

International Nuclear Information System (INIS)

Marwick, E.F.

1978-01-01

Disclosed is a nuclear reactor system which produces useful thermal power and breeds fissile isotopes wherein large spherical complex slugs containing fissile and fertile isotopes as well as vaporizing and tamping materials are exploded seriatim in a large containing chamber having walls protected from the effects of the explosion by about two thousand tons of slurry of fissile and fertile isotopes in molten alkali metal. The slug which is slightly sub-critical prior to its entry into the centroid portion of the chamber, then becomes slightly more than prompt-critical because of the near proximity of neutron-reflecting atoms and of fissioning atoms within the slurry. The slurry is heated by explosion of the slugs and serves as a working fluid for extraction of heat energy from the reactor. Explosive debris is precipitated from the slurry and used for the fabrication of new slugs

Laser Intertial Fusion Energy: Neutronic Design Aspects of a Hybrid Fusion-Fission Nuclear Energy System

Energy Technology Data Exchange (ETDEWEB)

Kramer, Kevin James [Univ. of California, Berkeley, CA (United States)

2010-04-08

This study investigates the neutronics design aspects of a hybrid fusion-fission energy system called the Laser Fusion-Fission Hybrid (LFFH). A LFFH combines current Laser Inertial Confinement fusion technology with that of advanced fission reactor technology to produce a system that eliminates many of the negative aspects of pure fusion or pure fission systems. When examining the LFFH energy mission, a significant portion of the United States and world energy production could be supplied by LFFH plants. The LFFH engine described utilizes a central fusion chamber surrounded by multiple layers of multiplying and moderating media. These layers, or blankets, include coolant plenums, a beryllium (Be) multiplier layer, a fertile fission blanket and a graphite-pebble reflector. Each layer is separated by perforated oxide dispersion strengthened (ODS) ferritic steel walls. The central fusion chamber is surrounded by an ODS ferritic steel first wall. The first wall is coated with 250-500 μm of tungsten to mitigate x-ray damage. The first wall is cooled by Li₁₇Pb₈₃ eutectic, chosen for its neutron multiplication and good heat transfer properties. The Li₁₇Pb₈₃ flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li₁₇Pb₈₃, separated from the Li₁₇Pb₈₃ by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF₂), chosen for beneficial neutronics and heat transfer properties. The use of flibe enables both fusion fuel production (tritium) and neutron moderation and multiplication for the fission blanket. A Be pebble (1 cm diameter) multiplier layer surrounds the coolant injection plenum and the coolant flows radially through perforated walls across the bed. Outside the Be layer, a fission fuel layer comprised of depleted uranium contained in Tristructural-isotropic (TRISO) fuel particles

Neutronics safety analysis in severe transients of the dual-cooled waste transmutation blanket for the FDS-I

International Nuclear Information System (INIS)

Zheng, S.; Wu, Y.

2006-01-01

The conceptual design of the fusion-driven subcritical system FDS-I with the multifunctional subcritical dual-cooled waste transmutation (DWT) blanket proposed as a middle step toward the final application of fusion energy was presented previously. Safety is a key concern for the innovative conceptual system. The potential safety characteristic is expected as one of the advantages of FDS-I for the transmutation and incineration of nuclear waste compared with the critical reactor. With the intent of evaluating the inherent and passive safety features of FDS-I, the reactivity coefficients (e.g. the coolant density/void effect and the Doppler feedback), which are similar to those in critical reactors, and the kinetics quantities (e.g. neutron generation time and β eff ), which are influenced by the external neutron source of the source-driven system, are calculated and analyzed. In addition, the specific transient scenarios for FDS-I, which show the effects of the source power perturbation on the safety parameters, is presented

Corrosion characteristics of an aqueous self-cooled fusion blanket

International Nuclear Information System (INIS)

Bogaerts, W.F.; Embrechts, M.J.; Steiner, D.; Deutsch, L.; Jackson, D.

1986-01-01

A novel aqueous self-cooled blanket concept (ASCB) has recently been proposed. This blanket concept, as applied to a MARS-like tandem mirror reactor, consists of disks of spiraling tubes of Zircaloy-4 housed in a structural container of vanadium alloy (V-15 Ti-5 Cr). The Zircaloy tubes are cooled by a mixture of light and heavy water with 9 g of LiOH per 100 cm 3 of water dissolved in the coolant. A major issue for the feasibility of the integrated blanket coil concept is the chemical compatibility of the coolant and Zircaloy. Initial corrosion tests have been undertaken in order to resolve this question. Results clearly show that successful alloy heats can be prepared, for which corrosion problems will probably not be the limiting factor of the ASCB design concept. As is quite well known from fission engineering studies, small variations in the alloy compositions or in the metallurgical structure may, however, be able to cause significant alterations in the oxidation or corrosion rates. Further tests will be necessary to resolve the remaining uncertainties and to determine the behavior of successful alloy heats in the presence of trace impurities in order to address the sensitivity to localized corrosion phenomena such as pitting, stress corrosion cracking, and intergranular attack

Accelerator-driven sub-critical target concept for transmutation of nuclear wastes

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Todosow, M.; Aronson, A.L.; Takahashi, H.; Geiger, M.J.

1991-01-01

A means of transmuting key long-lived nuclear wastes, primarily the minor actinides (Np, Am, Cm) and iodine, using a hybrid proton accelerator and sub-critical lattice, is proposed. By partitioning the components of the light water reactor (LWR) spent fuel and by transmuting key elements, such as the plutonium, the minor actinides, and a few of the long-lived fission products, some of the most significant challenges in building a waste repository can be substantially reduced. The proposed machine, based on the described PHOENIX Concept, would transmute the minor actinides and the iodine produced by 75 LWRs, and would generate usable electricity (beyond that required to run the large accelerator) of 850 MW e . 19 refs., 20 figs

Development of an engineering-scale nuclear test of a solid-breeder fusion-blanket concept

International Nuclear Information System (INIS)

Deis, G.A.; Bohn, T.S.; Hsu, P.Y.; Miller, L.G.; Scott, A.J.; Watts, K.D.; Welch, E.C.

1983-08-01

As part of the Phase I effort on Program Element-II (PE-II) of the Office of Fusion Energy/Argonne National Laboratory First Wall/Blanket/Shield Engineering Technology Program, a study has been performed to develop preconceptual hardware designs and preliminary test program descriptions for two fission-reactor-based tests of a water-cooled, solid-breeder fusion reactor blanket concept. First, a list of potentially acceptable reactor facilities is developed, based on a list of required reactor characteristics. From this set of facilities, two facilities are selected for study: the Oak Ridge Research Reactor (ORR) and the Power Burst Facility (PBF). A test which employs a cylindrical unit cell of a solid-breeder fusion reactor blanket, with pressurized-water cooling is designed for each facility. The test design is adjusted to the particular characteristics of each reactor. These two test designs are then compared on the basis of technical issues and cost. Both tests can satisfy the PE-II mission: blanket thermal hydraulic and thermomechanical issues. In addition, both reactors will produce prototypical tritium production rates and profiles and release characteristics with little or no additional modifications

An MCNP parametric study of George C. Laurence's subcritical pile experiment

International Nuclear Information System (INIS)

Dranga, R.; Blomeley, L.; Carrington, R.

2014-01-01

In the early 1940s at the National Research Council (NRC) Laboratories in Ottawa, Canada, Dr. George Laurence conducted several experiments to determine if a sustained nuclear fission chain reaction in a carbon-uranium arrangement (or 'pile') was possible. Although Dr. Laurence did not achieve criticality, these pioneering experiments marked a significant historical event in nuclear science, and they provided a valuable reference for subsequent experiments that led to the design of Canada's first heavy-water reactors at the Chalk River Nuclear Laboratories. This paper summarizes the results of a recent collaborative project between Atomic Energy of Canada Limited and the Deep River Science Academy undertaken to numerically explore the experiments carried out at the NRC Laboratories by Dr. Laurence, while teaching high school students about nuclear science and technology. In this study, a modern Monte Carlo reactor physics code, MCNP6, was utilized to identify and study the key parameters impacting the subcritical pile's neutron multiplication factor (e.g., moderation, geometry, material impurities) and quantify their effect on the extent of subcriticality. The findings presented constitute the first endeavour to model, using a current computational reactor physics tool, the seminal experiment that provided the foundation of Canada's nuclear science and technology program. (author)

Subcriticality monitoring method for reactor

International Nuclear Information System (INIS)

Ueda, Makoto.

1991-01-01

The present invention accurately monitors the reactor subcriticality and ensures the critical safety, irrespective of the presence or absence of artificial neutron sources. That is, when the subcriticality is monitored upon reactivity changing operation which causes reactivity change to the reactor during shutdown, neutron monitors are disposed at a plurality of monitoring positions. Then, neutron counting ratio before and after conducting the reactivity changing operation is determined. The subcriticality of the reactor is monitored by the ratio and the state of scattering of the ratio of neutron counting rate between each of the neutron monitors. With such procedures, signals of the neutron monitors are used, the characteristic that the change of the signals depend on the change of the neutron multiplication of the reactor core can be utilized whether artificial neutron sources (external neutron sources) are disposed or not. Accordingly, the subcriticality can be monitored more reliably. (I.S.)

Overview of Fusion-Fission Hybrid Reactor Design Study in China

International Nuclear Information System (INIS)

Huang Jinhua; Feng Kaiming; Deng Baiquan; Deng, P.Zh.; Zhang Guoshu; Hu Gang; He Kaihui; Wu Yican; Qiu Lijian; Huang Qunying; Xiao Bingjia; Liu Xiaoping; Chen Yixue; Kong, M.H.

2002-01-01

The motivation for developing fusion-fission hybrid reactors is discussed in the context of electricity power requirements by 2050 in China. A detailed conceptual design of the Fusion Experimental Breeder (FEB) was developed from 1986-1995. The FEB has a subignited tokamak fusion core with a major radius of 4.0 m, a fusion power of 145 MW, and a fusion energy gain Q of 3. Based on this, an engineering outline design study of the FEB, FEB-E, has been performed. This design study is a transition from conceptual to engineering design in this research. The main results beyond that given in the detailed conceptual design are included in this paper, namely, the design studies of the blanket, divertor, test blanket, and tritium and environment issues. In-depth analyses have been performed to support the design. Studies of related advanced concepts such as the waste transmutation blanket concept and the spherical tokamak core concept are also presented

Measurement of subcriticality by a pulsing α-method

International Nuclear Information System (INIS)

Jitarev, V.E.; Kachanov, V.M.; Kuzmin, A.N.

1999-01-01

The report presents results of a pulsing α-method [1] for determination of the WWER system subcriticality. The pulsing α-method permits to conduct measurements of system subcriticality in conditions of subcritical state and large neutron background. Therefore this method can be used for the control of a subcriticality of storehouses of a burn up nuclear fuel and stopped reactor. (Authors)

TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

Energy Technology Data Exchange (ETDEWEB)

Powers, Jeffrey James [Univ. of California, Berkeley, CA (United States)

2011-11-30

This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated

Hybrid reactors: Nuclear breeding or energy production?

International Nuclear Information System (INIS)

Piera, Mireia; Lafuente, Antonio; Abanades, Alberto; Martinez-Val, J.M.

2010-01-01

After reviewing the long-standing tradition on hybrid research, an assessment model is presented in order to characterize the hybrid performance under different objectives. In hybrids, neutron multiplication in the subcritical blanket plays a major role, not only for energy production and nuclear breeding, but also for tritium breeding, which is fundamental requirement in fusion-fission hybrids. All three objectives are better achieved with high values of the neutron multiplication factor (k-eff) with the obvious and fundamental limitation that it cannot reach criticality under any event, particularly, in the case of a loss of coolant accident. This limitation will be very important in the selection of the coolant. Some general considerations will be proposed, as guidelines for assessing the hybrid potential in a given scenario. Those guidelines point out that hybrids can be of great interest for the future of nuclear energy in a framework of Sustainable Development, because they can contribute to the efficient exploitation of nuclear fuels, with very high safety features. Additionally, a proposal is presented on a blanket specially suited for fusion-fission hybrids, although this reactor concept is still under review, and new work is needed for identifying the most suitable blanket composition, which can vary depending on the main objective of the hybrid.

Tritium chemistry in fission and fusion reactors

International Nuclear Information System (INIS)

Roth, E.; Masson, M.; Briec, M.

1986-09-01

We are interested in the behaviour of tritium inside the solids where it is generated both in the case of fission nuclear reactor fuel elements, and in that of blankets of future fusion reactor. In the first case it is desirable to be able to predict whether tritium will be found in the hulls or in the uranium oxide, and under what chemical form, in order to take appropriate steps for it's removal in reprocessing plants. In fusion reactors breeding large amounts of tritium and burning it in the plasma should be accomplished in as short a cycle as possible in order to limit inventories that are associated with huge activities. Mastering the chemistry of every step is therefore essential. Amounts generated are not of the same order of magnitude in the two cases studied. Ternary fissions produce about 66 10 13 Bq (18 000 Ci) per year of tritium in a 1000 MWe fission generator, i.e., about 1.8 10 10 Bq (0.5 Ci) per day per ton of fuel

ENHANCED ELECTROCHEMICAL PROCESSES IN SUBCRITICAL WATER

Energy Technology Data Exchange (ETDEWEB)

Steven B. Hawthorne

2000-07-01

This project involved designing and performing preliminary electrochemical experiments in subcritical water. An electrochemical cell with substantially better performance characteristics than presently available was designed, built, and tested successfully. The electrochemical conductivity of subcritical water increased substantially with temperature, e.g., conductivities increased by a factor of 120 when the temperature was increased from 25 to 250 C. Cyclic voltammograms obtained with platinum and nickel demonstrated that the voltage required to produce hydrogen and oxygen from water can be dropped by a factor of three in subcritical water compared to the voltages required at ambient temperatures. However, no enhancement in the degradation of 1,2-dichlorobenzene and the polychlorinated biphenyl 3,3',4,4'-tetrachlorobiphenyl was observed with applied potential in subcritical water.

Benchmarking criticality safety calculations with subcritical experiments

International Nuclear Information System (INIS)

Mihalczo, J.T.

1984-06-01

Calculation of the neutron multiplication factor at delayed criticality may be necessary for benchmarking calculations but it may not be sufficient. The use of subcritical experiments to benchmark criticality safety calculations could result in substantial savings in fuel material costs for experiments. In some cases subcritical configurations could be used to benchmark calculations where sufficient fuel to achieve delayed criticality is not available. By performing a variety of measurements with subcritical configurations, much detailed information can be obtained which can be compared directly with calculations. This paper discusses several measurements that can be performed with subcritical assemblies and presents examples that include comparisons between calculation and experiment where possible. Where not, examples from critical experiments have been used but the measurement methods could also be used for subcritical experiments

A neutron amplifier: prospects for reactor-based waste transmutation

International Nuclear Information System (INIS)

Blanovsky, A.

2004-01-01

A design concept and characteristics for an epithermal breeder controlled by variable feedback and external neutron source intensity are presented. By replacing the control rods with neutron sources, we could maintain good power distribution and perform radioactive waste burning in high flux subcritical reactors (HFSR) that have primary system size, power density and cost comparable to a pressurized water reactor (PWR). Another approach for actinide transmutation is a molten salt subcritical reactor proposed by Russian scientists. To increase neutron source intensity the HFSR is divided into two zones: a booster and a blanket with solid and liquid fuels. A neutron gate (absorber and moderator) imposed between two zones permits fast neutrons from the booster to flow to the blanket. Neutrons moving in the reverse direction are moderated and absorbed in the absorber zone. In the HFSR, neptunium-plutonium fuel is circulated in the booster and blanket, and americium-curium in the absorber zone and outer reflector. Use of a liquid actinide fuel permits transport of the delayed-neutron emitters from the blanket to the booster, where they can provide additional neutrons (source-dominated mode) or all the necessary excitation without an external neutron source (self-amplifying mode). With a blanket neutron multiplication gain of 20 and a booster gain of 50, an external neutron source rate of at least 10 15 n/s (0.7 MW D-T or 2.5 MW electron beam power) is needed to control the HFSR that produces 300 MWt. Most of the power could be generated in the blanket that burns about 100 kg of actinides a year. The analysis takes into consideration a wide range of HFSR design aspects including the wave model of observed relativistic phenomena, plant seismic diagnostics, fission electric cells (FEC) with a multistage collector (anode) and layered cathode. (author)

Calculation of the distribution of the escaping from a fissionable sample neutrons number when introducing one fission neutron in that sample

International Nuclear Information System (INIS)

Dorlet, J.

1991-01-01

A describing algorithm furnishes the probabilities of having exactly N escaping neutrons in the descent of one fission neutron, using the punctual reactor model. Calculations can be performed even for N-values greater than 1000. Numerical results show that discrete neutrons counting is unfit to obtain the N mean value, even for very far subcritical devices. That mean value is still very used in the existing theoretical studies, because its obvious correlation with the device effective multiplication coefficient. For that reason modelling coincidence neutrons counting is not suitable using statistical moments approach, but only using probabilities it selves [fr

Response distributions of 6LiF and 7LiF thermoluminescence dosimeters in lithium blanket assemblies

International Nuclear Information System (INIS)

Maekawa, Hiroshi; Kusano, Jyo-ichi; Seki, Yasushi

1976-11-01

Measurement of the radiation-heating rate distribution in the fusion blanket is as important as measurement of the fission-rate distribution in a fission reactor. To obtain the information of radiation heating, the response (integral glow value) distributions in pseudo-spherical lithium assemblies with and without a graphite reflector were measured with 6 LiF and 7 LiF TLD's. The measured responses are normalized to values per source neutron. Experimental error is about 35%, and the error in positions of TLD's is about +- 3 mm. The experimental results are compared with those of calculation using RADHEAT code system and ENDF/B-III data file. (auth.)

Transmutation blanket design for a Tokamak system

International Nuclear Information System (INIS)

Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria A. Fortini; Costa, Antonella L.

2011-01-01

Sub-critical advanced reactor with a D-T fusion neutron source based on Tokamak technology is an innovative type of nuclear system. Due to the high quantity of neutrons produced by fusion reactions, it could be well spent in the transmutation process of the transuranic elements. Nevertheless, to achieve a successful transmutation, it is necessary to know the neutron fluence along the radial axis and its characteristics. In this work, it evaluated the neutron flux and interaction frequency along the radial axis changing the material of the first wall. W-alloy, beryllium and the combination of both were studied and regions more suitable to transmutation were determined. The results demonstrated that the better zone to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements W-alloy/W-alloy and W-alloy/Beryllium would be able to hold the requirements of high fluence and hardening spectrum needed to transuranic transmutation. The system was simulated using the MCNP5 code, the ITER Final Design Report, 2001, and the FENDL/MC-2.1 nuclear data library. (author)

Remarks on the thorium cycle

International Nuclear Information System (INIS)

Teller, E.

1978-01-01

The use of thorium and neutrons to make 233 U would provide energy for many thousands of years. Thorium is more abundant than uranium and 233 U is the best fissile material for thermal neutron reactors. Four approaches to the use of thorium are worth developing: heavy water moderated reactors with conversion ratios greater than 0.9, such as modified CANDU with lower cost of separating D 2 O and 235 U; molten salt breeder reactors, from which fission products and excess fuel may be continuously removed; fusion-fission hybrids that produce adequate tritium and excess neutrons for sustenance and 233 U production in a subcritical thorium 233 U blanket; and by fission-initiated thermo-nuclear explosions in cavities in salt beds one mile below the earth's surface, yielding 233 U from the excess neutrons and thorium and decontaminated steam for power production. (author)

Swelling behavior detection of irradiated U-10Zr alloy fuel using indirect neutron radiography

Energy Technology Data Exchange (ETDEWEB)

Sun, Yong; Huo, He-yong; Wu, Yang [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Key Laboratory of Neutron Physics, Chinese Academy of Engineering Physics, Mianyang (China); Li, Jiangbo [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Zhou, Wei; Guo, Hai-bing [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Key Laboratory of Neutron Physics, Chinese Academy of Engineering Physics, Mianyang (China); Li, Hang, E-mail: lihang32@gmail.com [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Key Laboratory of Neutron Physics, Chinese Academy of Engineering Physics, Mianyang (China); Cao, Chao; Yin, Wei; Wang, Sheng; Liu, Bin; Feng, Qi-jie; Tang, Bin [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Key Laboratory of Neutron Physics, Chinese Academy of Engineering Physics, Mianyang (China)

2016-11-21

It is hopeful that fusion-fission hybrid energy system will become an effective approach to achieve long-term sustainable development of fission energy. U-10Zr alloy (which means the mass ratio of Zr is 10%) fuel is the key material of subcritical blanket for fusion-fission hybrid energy system which the irradiation performance need to be considered. Indirect neutron radiography is used to detect the irradiated U-10Zr alloy because of the high residual dose in this paper. Different burnup samples (0.1%, 0.3%, 0.5% and 0.7%) have been tested with a special indirect neutron radiography device at CMRR (China Mianyang Research Reactor). The resolution of the device is better than 50 µm and the quantitative analysis of swelling behaviors was carried out. The results show that the swelling behaviors relate well to burnup character which can be detected accurately by indirect neutron radiography.

Subcritical assemblies, use and their feasibility assessment

International Nuclear Information System (INIS)

Haroon, M.R.

1982-03-01

In developing countries, subcritical assemblies can be a useful tool for training and research in the field of nuclear technology with minimum cost. The historical development of subcritical assemblies and the reactor physics experiments which can be carried out using this facility are outlined. The different types of subcritical assemblies have been described and material requirements for each assembly have been pointed out. (author)

Monitoring of MNSR operation by measuring subcritical photoneutron flux

International Nuclear Information System (INIS)

Haddad, Kh.; Alsomel, N.

2011-01-01

Passive nondestructive assay methods are used to monitor the reactor's operation. It is required for nuclear regulatory, calculation validation and safeguards purposes. So, it plays a vital role in the safety and security of the nuclear plants. The possibility of MNSR operation monitoring by measuring the subcritical state photoneutron flux were investigated in this work. The photoneutron flux is induced by the fuels hard gamma radiation in the beryllium reflector. Theoretical formulation and experimental tests were performed. The results show that within a specified cooling time range, the photoneutron flux is induced by a single dominant hard gamma emitter such as 117 Cd (activation product) and 140 Ba ( 140 La fission product). This phenomenon was utilized to monitor the cooling time and the operation neutron flux during the last campaign. Thus a passive nondestructive assay method is proposed with regard to the reactor operation's monitoring.

Measurement of multiple α-modes at the Delphi subcritical assembly by neutron noise techniques

International Nuclear Information System (INIS)

Szieberth, Máté; Klujber, Gergely; Kloosterman, Jan Leen; Haas, Dick de

2015-01-01

Highlights: • Neutron noise measurements were performed at the Delphi subcritical assembly. • Bias in the fitted prompt decay constant was observed due to higher modes. • Spatial dependence of the higher mode was surveyed. • Effect of two different source distributions was investigated. • An estimation of the prompt decay constant is given for the Delphi. - Abstract: The paper presents the results and evaluations of a comprehensive set of neutron noise measurements on the Delphi subcritical assembly of the Delft University of Technology. The measurements investigated the effect of different source distributions (inherent spontaneous fission and 252 Cf) and the position of the detectors applied (both radially and vertically). The evaluation of the measured data has been performed by the variance-to-mean (VTM, Feynman-α), the autocorrelation (ACF, Rossi-α) and the cross-correlation (CCF) methods. The values obtained for the prompt decay constant show a strong bias, which depends both on the detector position and on the source distribution. This is due to the presence of higher modes in the system. It has been observed that the α value fitted is higher when the detector is close to the boundary of the core or to the 252 Cf point-source. The higher alpha-modes have also been observed by fitting functions describing two alpha-modes. The successful set of measurements also provides a good basis for further theoretical investigations including the Monte Carlo simulation of the noise measurements and the calculation of the alpha-modes in the Delphi subcritical assembly

Preconceptual design of a packed fluidized bed blanket for a fission suppressed thorium-fueled CTHR

International Nuclear Information System (INIS)

Chi, J.W.H.; Karbowski, J.S.; Chapin, D.L.

1981-01-01

This paper describes a thorium-fueled PFB blanket concept for a Commercial Tokamak Hybrid Reactor. A preliminary mechanical concept is presented and the results of neutronics, thermal-hydraulics and economics analyses are discussed. Futher work needed to design and advance the concept is recommended

Progress on the conceptual design of a mirror hybrid fusion--fission reactor

International Nuclear Information System (INIS)

Moir, R.W.; Lee, J.D.; Burleigh, R.J.

1975-01-01

A conceptual design study was made of a fusion-fission reactor for the purpose of producing fissile material and electricity. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and is sustained by neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and helium cooled. It was shown conceptually how the reactor might be built using essentially present-day technology and how the uranium-bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel

Design project of fast subcritical system 'Mala Lasta'

International Nuclear Information System (INIS)

Milosevic, M.; Stefanovic, D.; Popovic, D.; Pesic, M.; Zavaljevski, N.; Nikolic, D.; Arsenovic, M.

1988-10-01

This report contains two parts. Part one covers the objective and fundamental elements for the choice of fast subcritical system 'Mala Lasta', review of the existing fast subcritical assemblies, and a description of the available domestic computer codes applied for calculating neutron reactor parameters. Comparison of results obtained by these codes for a number of existing subcritical assemblies was used for the choice of the design project described in part two of this report. It contains detailed description of the operating parameters of the chosen subcritical system based on the obtained calculated parameters

The Los Alamos accelerator-driven transmutation of nuclear waste (ATW) concept development of the ATW target/blanket system

International Nuclear Information System (INIS)

Venneri, F.; Williamson, M.A.; Ning, L.

1997-01-01

In the past several years, the Los Alamos ADTT program has conducted studies of an innovative technology for solving the nuclear waste problem and building a new generation of safer and non-proliferant nuclear power plants. The ATW concept destroys higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. In this paper special attention is given to the basic design of the ATW Molten Salt concept and the safety perspective. 40 refs., 11 figs

Cascade energy amplifier

International Nuclear Information System (INIS)

Barzilov, A.P.; Gulevich, A.V.; Kukharchuk, O.F.

2000-01-01

The technical problem of long-life fission product and minor actinide incineration and production of plutonium fuel in the prospective nuclear systems will arise at significant scales of nuclear power industry development. Subcritical nuclear reactors driven by extemal neutron sources (energy amplifiers) are considered as incinerators of toxicity of complete nuclear industry. In the frames of this concept, the subcritical reactor part consisting of two coupled blanket regions (inner fast neutron spectrum core and outer thermal core) driven by extemal neutron source is discussed. Two types of source are studied: spallation target and 14-MeV fusion bum of micropellets. Liquid metal Pb-Bi is considered as target material and coolant of inner fast core. Thermal core is a heavy-water subcritical reactor of the Candu-type. The fast core is protected from thermal neutrons influence with the boron shield. All reactor technologies used in this concept are tested during years of operation and commercially available. Thus, the cascade energy amplifiers have a set of advantages in comparison with traditional concepts: in energy production, in transmutation efficiency, and in economics. (authors)

Deep subcritical levels measurements dependents upon kinetic distortion factors

International Nuclear Information System (INIS)

Pan Shibiao; Li Xiang; Fu Guo'en; Huang Liyuan; Mu Keliang

2013-01-01

The measurement of deep subcritical levels, with the increase of subcriticality, showed that the results impact on the kinetic distortion effect, along with neutron flux strongly deteriorated. Using the diffusion theory, calculations have been carried out to quantify the kinetic distortion correction factors in subcritical systems, and these indicate that epithermal neutron distributions are strongly affected by kinetic distortion. Subcriticality measurements in four different rod-state combination at the zero power device was carried out. The test data analysis shows that, with increasing subcriticality, kinetic distortion effect correction factor gradually increases from 1.052 to 1.065, corresponding reactive correction amount of 0.78β eff ∼ 3.01β eff . Thus, it is necessary to consider the kinetic distortion effect in the deep subcritical reactivity measurements. (authors)

Consultancy to review and finalize the IAEA publication 'Compendium on the use of fusion/fission hybrids for the utilization and transmutation of actinides and long-lived fission products'. Working material

International Nuclear Information System (INIS)

2004-01-01

In addition to the traditional fission reactor research, fusion R and D activities are becoming of interest also to nuclear fission power development. There is renewed interest in utilizing fusion neutrons, Heavy Liquid Metals, and molten salts for innovative systems (energy production and transmutation). Indeed, for nuclear power development to become sustainable as a long-term energy option, innovative fuel cycle and reactor technologies will have to be developed to solve the problems of resource utilization and long-lived radioactive waste management. In this context Member States clearly expressed the need for comparative assessments of various transmutation reactors. Both the fusion and fission communities are currently investigating the potential of innovative reactor and fuel cycle strategies that include a fusion/fission system. The attention is mainly focused on substantiating the potential advantages of such systems: utilization and transmutation of actinides and long-lived fission products, intrinsic safety features, enhanced proliferation resistance, and fuel breeding capabilities. An important aspect of the ongoing activities is the comparison with the accelerator driven subcritical system (spallation neutron source), which is the other main option for producing excess neutrons. Apart from comparative assessments, knowledge preservation is another subject of interest to the Member States: the goal, applied to fusion/fission systems, is to review the status of, and to produce a 'compendium' of past and present achievements in this area

An MCNP parametric study of George C. Laurence's subcritical pile experiment

Energy Technology Data Exchange (ETDEWEB)

Dranga, R.; Blomeley, L., E-mail: ruxandra.dranga@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carrington, R. [McGill Univ., Dept. of Mathematics and Statistics, Montreal, Quebec (Canada)

2014-12-01

In the early 1940s at the National Research Council (NRC) Laboratories in Ottawa, Canada, Dr. George Laurence conducted several experiments to determine if a sustained nuclear fission chain reaction in a carbon-uranium arrangement (or 'pile') was possible. Although Dr. Laurence did not achieve criticality, these pioneering experiments marked a significant historical event in nuclear science, and they provided a valuable reference for subsequent experiments that led to the design of Canada's first heavy-water reactors at the Chalk River Nuclear Laboratories. This paper summarizes the results of a recent collaborative project between Atomic Energy of Canada Limited and the Deep River Science Academy undertaken to numerically explore the experiments carried out at the NRC Laboratories by Dr. Laurence, while teaching high school students about nuclear science and technology. In this study, a modern Monte Carlo reactor physics code, MCNP6, was utilized to identify and study the key parameters impacting the subcritical pile's neutron multiplication factor (e.g., moderation, geometry, material impurities) and quantify their effect on the extent of subcriticality. The findings presented constitute the first endeavour to model, using a current computational reactor physics tool, the seminal experiment that provided the foundation of Canada's nuclear science and technology program. (author)

Major features of a mirror fusion--fast fission hybrid reactor

International Nuclear Information System (INIS)

Moir, R.W.; Lee, J.D.; Burleigh, R.J.

1974-01-01

A conceptual design was made of a fusion-fission reactor. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and sustained by hot neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and is cooled by helium. It was shown how the reactor can be built using essentially present day construction technology and how the uranium bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel of which approximately 1200 kg of plutonium are produced each year along with the approximately 750 MW of electricity. (U.S.)

Designing a mini subcritical nuclear reactor

International Nuclear Information System (INIS)

Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M.

2015-10-01

In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of 239 PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of 239 PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k e -f f , the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k eff , the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons 239 PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k eff was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k eff of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five cylinders not met. (Author)

Doubling and half-life times for a combined fission-fusion system

Energy Technology Data Exchange (ETDEWEB)

Pantis, G

1982-01-01

The long term fuel dynamics for a fission symbiotic system is examined under the assumption of discontinuous loading and offloading conditions. It is found that the breeding capacities of the core and the blankets identify several distinct fuel cycles. By numerical test and a specific comparison it is shown that doubling times and half-lives can differ by as much as 10% from those predicted by conventional methods.

The influence of external source intensity in accelerator/target/blanket system on conversion ratio and fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Kochurov, B.P. [Institute for Theoretical and Experimental Physics, Moscow (Russian Federation)

1995-10-01

The analysis of neutron balance relation for a subcritical system with external source shows that a high ratio of neutron utilization (conversion ratio, breeding ratio) much exceeding similar values for nuclear reactors (both thermal or fast spectrum) is reachable in accelerator/target/blanket system with high external neutron source intensity. An accelerator/target/blanket systems with thermal power in blanket about 1850 Mwt and operating during 30 years have been investigated. Continual feed up by plutonium (fissile material) and Tc-99 (transmuted material) was assumed. Accelerator beam intensity differed 6.3 times (16 mA - Case 1, and 100 mA-Case 2). Conversion ratio (CR) was defined as the ratio of Tc-99 nuclei transmuted to the number of Pu nuclei consumed. High value of conversion ratio considerably exceeding 1 (CR=1.66) was obtained in the system with high source intensity as compared with low source system (CR=0.77). Net output of electric power of high source intensity system is about twice lower due to consumption of electric power for accelerator feed up. The loss of energy for Tc-99 transmutation is estimated as 40 Mev(el)/nuclei. Yet high conversion ratio (or breeding ratio) achievable in electronuclear installations with high intensity of external source can effectively be used to close fuel cycle (including incineration of wastes) or to develop growing nuclear power production system.

Present status of irradiation tests on tritium breeding blanket for fusion reactor

International Nuclear Information System (INIS)

Futamura, Yoshiaki; Sagawa, Hisashi; Shimakawa, Satoshi; Tsuchiya, Kunihiko; Kuroda, Toshimasa; Kawamura, Hiroshi.

1994-01-01

To develop a tritium breeding blanket for a fusion reactor, irradiation tests in fission reactors are indispensable for obtaining data on irradiation effects on materials, and neutronics/thermal characteristics and tritium production/recovery performance of the blanket. Various irradiation tests have been conducted in the world, especially to investigate tritium release characteristics from tritium breeding and neutron multiplier materials, and materials integrity under irradiation. In Japan, VOM experiments at JRR-2 for ceramic breeders and experiments at JMTR for ceramic breeders and beryllium as a neutron multiplier have been performed. Several universities have also investigated ceramic breeders. In the EC, the EXOTIC experiments at HFR in the Netherlands and the SIBELIUS, the LILA, the LISA and the MOZART experiments for ceramic breeders have carried out. In Canada, NRU has been used for the CRITIC experiments. The TRIO experiments at ORR(ORNL), experiments at RTNS-II, FUBR and ATR have been conducted in the USA. The last two are experiments with high neutron fluence aiming at investigating materials integrity under irradiation. The BEATRIX-I and -II experiments have proceeded under international collaboration of Japan, Canada, the EC and the USA. This report shows the present status of these irradiation tests following a review of the blanket design in the ITER CDA(Conceptual Design Activity). (author)

Materials for breeding blankets

International Nuclear Information System (INIS)

Mattas, R.F.; Billone, M.C.

1995-09-01

There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as Primary Blanket Materials, which have the greatest influence in determining the overall design and performance, and Secondary Blanket Materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified

Materials for breeding blankets

International Nuclear Information System (INIS)

Mattas, R.F.; Billone, M.C.

1996-01-01

There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as primary blanket materials, which have the greatest influence in determining the overall design and performance, and secondary blanket materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified. (orig.)

Experimental investigations of the accelerator-driven transmutation technologies at the subcritical facility ''Yalina''

International Nuclear Information System (INIS)

Chigrinov, S.E.; Kiyavitskaya, H.I.; Serafimovich, I.G.; Rakhno, I.L.; Rutkovskaia, Ch.K.; Fokov, Y.; Khilmanovich, A.M.; Marstinkevich, B.A.; Bournos, V.V.; Korneev, S.V.; Mazanik, S.E.; Kulikovskaya, A.V.; Korbut, T.P.; Voropaj, N.K.; Zhouk, I.V.; Kievec, M.K.

2002-01-01

The investigations on accelerator-driven transmutation technologies (ADTT) focus on the reduction of the amount of long-lived wastes and the physics of a subcritical system driven with an external neutron source. This paper presents the experimental facility 'Yalina' which was designed and created at the Radiation Physics and Chemistry Problems Institute of the National Academy of Sciences of Belarus in the framework of the ISTC project no. B-070 to study the peculiarities of ADTT in thermal spectrum. A detailed description of the assembly, neutron generator and a preliminary analysis of some calculated and experimental data (multiplication factor, neutron flux density distribution in the assembly, transmutation rates of some long-lived fission products and minor actinides) are presented. (authors)

Steady squares and hexagons on a subcritical ramp

International Nuclear Information System (INIS)

Hoyle, R.B.

1995-01-01

Steady squares and hexagons on a subcritical ramp are studied, both analytically and numerically, within the framework of the lowest-order amplitude equations. On the subcritical ramp, the external stress or control parameter varies continuously in space from subcritical to supercritical values. At the subcritical end of the ramp, pattern formation is suppressed, and patterns fade away into the conduction solution. It is shown that three-dimensional patterns may change shape on a subcritical ramp. A square pattern becomes a pattern of rolls as it fades, with the roll axes aligned in the direction orthogonal to that in which the control parameter varies. Hexagons in systems with horizontal midplane symmetry become a pattern of rectangles before reaching the conduction solution. There is a suggestion that hexagons in systems which lack this symmetry might fade away through a roll pattern. Numerical simulations are used to illustrate these phenomena

System model for analysis of the mirror fusion-fission reactor

International Nuclear Information System (INIS)

Bender, D.J.; Carlson, G.A.

1977-01-01

This report describes a system model for the mirror fusion-fission reactor. In this model we include a reactor description as well as analyses of capital cost and blanket fuel management. In addition, we provide an economic analysis evaluating the cost of producing the two hybrid products, fissile fuel and electricity. We also furnish the results of a limited parametric analysis of the modeled reactor, illustrating the technological and economic implications of varying some important reactor design parameters

Material synergism fusion-fission

International Nuclear Information System (INIS)

Sankara Rao, K.B.; Raj, B.; Cook, I.; Kohyama, A.; Dudarev, S.

2007-01-01

In fission and fusion reactors the common features such as operating temperatures and neutron exposures will have the greatest impact on materials performance and component lifetimes. Developing fast neutron irradiation resisting materials is a common issue for both fission and fusion reactors. The high neutron flux levels in both these systems lead to unique materials problems like void swelling, irradiation creep and helium embitterment. Both fission and fusion rely on ferritic-martensitic steels based on 9%Cr compositions for achieving the highest swelling resistance but their creep strength sharply decreases above ∝ 823K. The use of oxide dispersion strengthened (ODS) alloys is envisaged to increase the operating temperature of blanket systems in the fusion reactors and fuel clad tubes in fast breeder reactors. In view of high operating temperatures, cyclic and steady load conditions and the long service life, properties like creep, low cycle fatigue,fracture toughness and creepfatigue interaction are major considerations in the selection of structural materials and design of components for fission and fusion reactors. Currently, materials selection for fusion systems has to be based upon incomplete experimental database on mechanical properties. The usage of fairly well developed databases, in fission programmes on similar materials, is of great help in the initial design of fusion reactor components. Significant opportunities exist for sharing information on technology of irradiation testing, specimen miniaturization, advanced methods of property measurement, safe windows for metal forming, and development of common materials property data base system. Both fusion and fission programs are being directed to development of clean steels with very low trace and tramp elements, characterization of microstructure and phase stability under irradiation, assessment of irradiation creep and swelling behaviour, studies on compatibility with helium and developing

Large subcriticality measurement by pulsed neutron method

International Nuclear Information System (INIS)

Yamane, Y.; Yoshida, A.; Nishina, K.; Kobayashi, K.; Kanda, K.

1985-01-01

To establish the method determining large subcriticalities in the field of nuclear criticality safety, the authors performed pulsed neutron experiments using the Kyoto University Critical Assembly (KUCA) at Research Reactor Institute, Kyoto University and the Cockcroft-Walton type accelerator attached to the assembly. The area-ratio method proposed by Sjoestrand was employed to evaluate subcriticalities from neutron decay curves measured. This method has the shortcomings that the neutron component due to a decay of delayed neutrons remarkably decreases as the subcriticality of an objective increases. To overcome the shortcoming, the authors increased the frequency of pulsed neutron generation. The integral-version of the area-ratio method proposed by Kosaly and Fisher was employed in addition in order to remove a contamination of spatial higher modes from the decay curve. The latter becomes significant as subcriticality increases. The largest subcriticality determined in the present experiments was 125.4 dollars, which was equal to 0.5111 in a multiplication factor. The calculational values evaluated by the computer code KENO-IV with 137 energy groups based on the Monte Carlo method agreed well with those experimental values

ITER blanket designs

International Nuclear Information System (INIS)

Gohar, Y.; Parker, R.; Rebut, P.H.

1995-01-01

The ITER first wall, blanket, and shield system is being designed to handle 1.5±0.3 GW of fusion power and 3 MWa m -2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design. (orig.)

ITER convertible blanket evaluation

International Nuclear Information System (INIS)

Wong, C.P.C.; Cheng, E.

1995-01-01

Proposed International Thermonuclear Experimental Reactor (ITER) convertible blankets were reviewed. Key design difficulties were identified. A new particle filter concept is introduced and key performance parameters estimated. Results show that this particle filter concept can satisfy all of the convertible blanket design requirements except the generic issue of Be blanket lifetime. If the convertible blanket is an acceptable approach for ITER operation, this particle filter option should be a strong candidate

Burnup studies of the subcritical fusion-driven in-zinerator

International Nuclear Information System (INIS)

Persson, C. M.; Gudowski, W.; Venneri, F.

2007-01-01

A fusion-driven subcritical core, 'In-Zinerator', has been proposed for nuclear waste transmutation [1]. In this concept, a powerful Z-pinch neutron source will produce pulses of 14 MeV neutrons that multiply in a surrounding subcritical core consisting of spent fuel from the LWR fuel cycle or from deep burn high temperature reactors. The proposed design has pulse frequency 0.1 Hz and a thermal power of 3 GWth. The Z-pinch fusion experiment is located at Sandia Laboratories, USA, and can today fire once a day. However, investigations have been made how to increase the frequency to several fires per minute. Each fire yields 300 MJ corresponding to 1020 neutrons per pulse. The source chamber will in the In-Zinerator concept be surrounded by spent fuel to reach an effective multiplication factor, k e ff, of 0.97. The core will be cooled by liquid lead. In this paper, the burnup of different fuel compositions in the In-Zinerator will be studied as function of initial k e ff. The Monte Carlo based continuous energy burnup code MCB [2][3]will be used. References: [1] B.B. Cipiti, Fusion Transmutation of Waste and the Role of the In-Zinerator in the Nuclear Fuel Cycle, Sandia Report SAND2006-3522, Sandia National Laboratories, USA, 2006. [2] J. Cetnar, J Wallenius and W Gudowski, MCB: A continuous energy Monte-Carlo burnup simulation code, Actinide and fission product partitioning and transmutation, Proc. of the Fifth Int. Information Exchange Meeting, Mol, Belgium, 25-27 November 1998, 523, OECD/NEA, 1998. [3] http://www.nea.fr/abs/html/nea-1643.html

The requirements for processing tritium recovered from liquid lithium blankets: The blanket interface

International Nuclear Information System (INIS)

Clemmer, R.G.; Finn, P.A.; Greenwood, L.R.; Grimm, T.L.; Sze, D.K.; Bartlit, J.R.; Anderson, J.L.; Yoshida, H.; Naruse.

1988-03-01

We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs

Blanket testing in NET

International Nuclear Information System (INIS)

Chazalon, M.; Daenner, W.; Libin, B.

1989-01-01

The testing stages in NET for the performance assessment of the various breeding blanket concepts developed at the present time in Europe for DEMO (LiPb and ceramic blankets) and the requirements upon NET to perform these tests are reviewed. Typical locations available in NET for blanket testing are the central outboard segments and the horizontal ports of in-vessel sectors. These test positions will be connectable with external test loops. The number of test loops (helium, water, liquid metal) will be such that each major class of blankets can be tested in NET. The test positions, the boundary conditions and the external test loops are identified and the requirements for test blankets are summarized (author). 6

Pulse superimposition calculational methodology for estimating the subcriticality level of nuclear fuel assemblies

International Nuclear Information System (INIS)

Talamo, Alberto; Gohar, Y.; Rabiti, C.; Aliberti, G.; Kondev, F.; Smith, D.; Zhong, Z.; Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.; Serafimovich, I.

2009-01-01

One of the most reliable experimental methods for measuring the subcriticality level of a nuclear fuel assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology simulating the Sjoestrand method, which allows comparing the experimental and analytical reaction rates and the obtained subcriticality levels. In this methodology, the reaction rate is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the delayed fission neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction rate is vanished. The obtained reaction rate is then superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The analytical results of this new calculation methodology have shown an excellent agreement with the experimental data available from the YALINA-Booster facility of Belarus. This methodology can be used to calculate Bell and Glasstone spatial correction factor.

Pulse superimposition calculational methodology for estimating the subcriticality level of nuclear fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: atalamo@anl.gov; Gohar, Y. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Rabiti, C. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, ID 83403 (United States); Aliberti, G.; Kondev, F.; Smith, D.; Zhong, Z. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.; Serafimovich, I. [Joint Institute for Power and Nuclear Research-Sosny, National Academy of Sciences (Belarus)

2009-07-21

One of the most reliable experimental methods for measuring the subcriticality level of a nuclear fuel assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology simulating the Sjoestrand method, which allows comparing the experimental and analytical reaction rates and the obtained subcriticality levels. In this methodology, the reaction rate is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the delayed fission neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction rate is vanished. The obtained reaction rate is then superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The analytical results of this new calculation methodology have shown an excellent agreement with the experimental data available from the YALINA-Booster facility of Belarus. This methodology can be used to calculate Bell and Glasstone spatial correction factor.

Sub-criticality monitoring for ADTR trademark control

International Nuclear Information System (INIS)

Ashworth, Roger

2014-01-01

Following the debut of the Accelerator Driven Thorium Reactor (ADTR trademark) Power Station at ENC 2010 in Barcelona, thorium as a reactor fuel has gained increasing support. The ADTR trademark concept reactor introduced the combination of an accelerator driven system (ADS) with traditional control rod technology, to provide a very high gain novel sub-critical ADS reactor design. The high gain of the system, while significantly reducing the demands on the accelerator design, pushes up operational sub-criticality (k eff ) closer to unity. In this paper we review this design and the progress made since ENC 2010. We compare 2 different methods of measuring the sub-critical neutron multiplication factor as the fuel cycle develops. The paper discusses the most recent work on k eff measurement and the interesting relationship between neutron flux, accelerator current and fuel temperature when using beam pulse methods to determine operational sub-criticality, of which a European patent is being granted. (orig.)

Estimation of the sub-criticality of the sodium-cooled fast reactor Monju using the modified neutron source multiplication method

International Nuclear Information System (INIS)

Truchet, G.; Van Rooijen, W. F. G.; Shimazu, Y.; Yamaguchi, K.

2012-01-01

The Modified Neutron Source Method (MNSM) is applied to the Monju reactor. This static method to estimate sub-criticality has already given good results on commercial Pressurized Water Reactors. The MNSM consists both in the extraction of the fundamental mode seen by a detector to avoid the effect of higher modes near sources, and the correction of flux distortion effects due to control rod movement. Among Monju's particularities that have a big influence on MNSM factors are: the presence of two californium sources and the position of the detector which is located far from the core outside of the reactor vessel. The importance of spontaneous fission and (α, n) reactions which have increased during the shutdown period of 15 years will also be discussed. The relative position of detectors and sources deeply affect the correction factors in some regions. In order to evaluate the detector count rate, an analytical propagation has been conducted from the reactor vessel. For two subcritical states, an estimation of the reactivity has been made and compared to experimental data obtained in the restart experiments at Monju (2010). (authors)

Fusion reactor blanket-main design aspects

International Nuclear Information System (INIS)

Strebkov, Yu.; Sidorov, A.; Danilov, I.

1994-01-01

The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm 2 and 7 MWa/m 2 accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket

Limitations on blanket performance

International Nuclear Information System (INIS)

Malang, S.

1999-01-01

The limitations on the performance of breeding blankets in a fusion power plant are evaluated. The breeding blankets will be key components of a plant and their limitations with regard to power density, thermal efficiency and lifetime could determine to a large degree the attractiveness of a power plant. The performance of two rather well known blanket concepts under development in the frame of the European Blanket Programme is assessed and their limitations are compared with more advanced (and more speculative) concepts. An important issue is the question of which material (structure, breeder, multiplier, coatings) will limit the performance and what improvement would be possible with a 'better' structural material. This evaluation is based on the premise that the performance of the power plant will be limited by the blankets (including first wall) and not by other components, e.g. divertors, or the plasma itself. However, the justness of this premise remains to be seen. It is shown that the different blanket concepts cover a large range of allowable power densities and achievable thermal efficiencies, and it is concluded that there is a high incentive to go for better performance in spite of possibly higher blanket cost. However, such high performance blankets are usually based on materials and technologies not yet developed and there is a rather high risk that the development could fail. Therefore, it is explained that a part of the development effort should be devoted to concepts where the materials and technologies are more or less in hand in order to ensure that blankets for a DEMO reactor can be developed and tested in a given time frame. (orig.)

Fusion blanket design and optimization techniques

International Nuclear Information System (INIS)

Gohar, Y.

2005-01-01

In fusion reactors, the blanket design and its characteristics have a major impact on the reactor performance, size, and economics. The selection and arrangement of the blanket materials, dimensions of the different blanket zones, and different requirements of the selected materials for a satisfactory performance are the main parameters, which define the blanket performance. These parameters translate to a large number of variables and design constraints, which need to be simultaneously considered in the blanket design process. This represents a major design challenge because of the lack of a comprehensive design tool capable of considering all these variables to define the optimum blanket design and satisfying all the design constraints for the adopted figure of merit and the blanket design criteria. The blanket design techniques of the First Wall/Blanket/Shield Design and Optimization System (BSDOS) have been developed to overcome this difficulty and to provide the state-of-the-art techniques and tools for performing blanket design and analysis. This report describes some of the BSDOS techniques and demonstrates its use. In addition, the use of the optimization technique of the BSDOS can result in a significant blanket performance enhancement and cost saving for the reactor design under consideration. In this report, examples are presented, which utilize an earlier version of the ITER solid breeder blanket design and a high power density self-cooled lithium blanket design for demonstrating some of the BSDOS blanket design techniques

Neutronics analysis of minor actinides transmutation in a fusion-driven subcritical system

International Nuclear Information System (INIS)

Yang, Chao; Cao, Liangzhi; Wu, Hongchun; Zheng, Youqi; Zu, Tiejun

2013-01-01

Highlights: • A fusion fission hybrid system for MA transmutation is proposed. • The analysis of neutronics effects on the transmutation is performed. • The transmutation rate of MA reaches 86.5% by 25 times of recycling. -- Abstract: The minor actinides (MAs) transmutation in a fusion-driven subcritical system is analyzed in this paper. The subcritical reactor is driven by a tokamak D-T fusion device with relatively easily achieved plasma parameters and tokamak technologies. The MAs discharged from the light water reactor (LWR) are loaded in transmutation zone. Sodium is used as the coolant. The mass percentage of the reprocessed plutonium (Pu) in the fuel is raised from 0 to 48% and stepped by 12% to determine its effect on the MAs transmutation. The lesser the Pu is loaded, the larger the MAs transmutation rate is, but the smaller the energy multiplication factor is. The neutronics analysis of two loading patterns is performed and compared. The loading pattern where the mass percentage of Pu in two regions is 15% and 32.9% respectively is conducive to the improvement of the transmutation fraction within the limits of burn-up. The final transmutation fraction of MAs can reach 17.8% after five years of irradiation. The multiple recycling is investigated. The transmutation fraction of MAs can reach about 61.8% after six times of recycling, and goes up to about 86.5% after 25

Performance test of micro-fission chambers for in-vessel neutron monitoring of ITER

International Nuclear Information System (INIS)

Yamauchi, Michinori; Nishitani, Takeo; Ochiai, Kentaro; Morimoto, Yuichi; Hori, Jun-ichi; Ebisawa, Katsuyuki; Kasai, Satoshi

2002-03-01

A micro-fission chamber with 12 mg UO 2 and a dummy chamber without uranium were fabricated and the performance was tested. They are designed to be installed inside the vacuum vessel of the compact ITER (ITER-FEAT) for neutron monitoring. The vacuum leak rate of the dummy chamber with MI cable, resistances of chambers between central conductor and outer sheath, and mechanical strength up to 50G acceleration were confirmed to meet the design criteria. The gamma-ray sensitivity was measured for the dummy chamber with the 60 Co gamma-ray irradiation facility at JAERI Takasaki. The output signals for gamma-rays in Campbelling mode were estimated to be less than 0.1% of those by neutrons at the location behind the blanket module in ITER-FEAT. The detector response for 14 MeV neutrons was investigated with the FNS facility. Excellent linearity between count rates, square of Campbelling voltage and neutron fluxes was confirmed in the temperature range from 20degC (room) to 250degC. However, a positive dependence of 14 MeV neutron count rates on temperature was observed, which might be caused by the increase in the pulse height with temperature rise. Effects of a change of surrounding materials were evaluated by the sensitivity measurements of the micro-fission chamber inserted into the shielding blanket mock-up. The sensitivity was enhanced by slow-downed neutrons, which agreed with the calculation result by MCNP-4C code. As a result, it was concluded that the developed micro-fission chamber is applicable for ITER-FEAT. (author)

Systems Modeling For The Laser Fusion-Fission Energy (LIFE) Power Plant

International Nuclear Information System (INIS)

Meier, W.R.; Abbott, R.; Beach, R.; Blink, J.; Caird, J.; Erlandson, A.; Farmer, J.; Halsey, W.; Ladran, T.; Latkowski, J.; MacIntyre, A.; Miles, R.; Storm, E.

2008-01-01

A systems model has been developed for the Laser Inertial Fusion-Fission Energy (LIFE) power plant. It combines cost-performance scaling models for the major subsystems of the plant including the laser, inertial fusion target factory, engine (i.e., the chamber including the fission and tritium breeding blankets), energy conversion systems and balance of plant. The LIFE plant model is being used to evaluate design trade-offs and to identify high-leverage R and D. At this point, we are focused more on doing self consistent design trades and optimization as opposed to trying to predict a cost of electricity with a high degree of certainty. Key results show the advantage of large scale (>1000 MWe) plants and the importance of minimizing the cost of diodes and balance of plant cost

ITER shielding blanket

Energy Technology Data Exchange (ETDEWEB)

Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

1995-03-01

A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

Fusion fuel blanket technology

International Nuclear Information System (INIS)

Hastings, I.J.; Gierszewski, P.

1987-05-01

The fusion blanket surrounds the burning hydrogen core of a fusion reactor. It is in this blanket that most of the energy released by the nuclear fusion of deuterium-tritium is converted into useful product, and where tritium fuel is produced to enable further operation of the reactor. As fusion research turns from present short-pulse physics experiments to long-burn engineering tests in the 1990's, energy removal and tritium production capabilities become important. This technology will involve new materials, conditions and processes with applications both to fusion and beyond. In this paper, we introduce features of proposed blanket designs and update and status of international research. In focusing on the Canadian blanket technology program, we discuss the aqueous lithium salt blanket concept, and the in-reactor tritium recovery test program

Study of DD versus DT fusion fuel cycles for different fusion-fission hybrid energy systems

International Nuclear Information System (INIS)

Gohar, Y.; Baker, C.C.

1981-01-01

A study was performed to investigate the characteristics of an energy system to produce fissile fuel for fission reactors. DD and DT fusion reactors were examined in this study with either a thorium or uranium blanket for each fusion reactor. Various fuel cycles were examined for light-water reactors including the denatured fuel cycles (which may offer proliferation resistance compared to other fuel cycles); these fuel cycles include a uranium fuel cycle with 239 Pu makeup, a thorium fuel cycle with 239 Pu makeup, a denatured uranium fuel cycle with 233 U makeup, and a denatured thorium fuel cycle with 233 U makeup. Four different blankets were considered for this study. The first two blankets have a tritium breeding capability for DT reactors. Lithium oxide (Li 2 O) was used for tritium breeding due to its high lithium density and high temperature capability; however, the use of Li 2 O may result in higher tritium inventories compared to other solid breeders

Research on transmutation and accelerator-driven systems at the Forschungszentrum Karlsruhe

International Nuclear Information System (INIS)

Knebel, J.U.; Heusener, G.

2000-01-01

Transmutation is considered a promising technology worldwide for significantly reducing the amount and, thereby, the long-term radiotoxicity of high active waste (HAW) produced by the operation of nuclear power plants such as light water reactors (LWR). The maximum reduction of radiotoxicity could be by a factor of about 100. Transmutation is thus an alternative to the direct deposition of large volumes of highly radioactive waste. Transmutation presents the possibility of closing the fuel cycle including the minor actinides. Plutonium, minor actinides and long-lived fission products can be transmuted in a so called Accelerator Driven Sub-critical System (ADS), which consists of an accelerator, a target module and a subcritical blanket. This paper describes the work performed at Forschungszentrum Karlsruhe which is critically evaluating an ADS mainly with respect to its potential for transmuting minor actinides, to its feasibility and to safety aspects. The work is being done in the area of core design, neutronics, safety, system analyses, materials and corrosion. (orig.) [de

Development of blanket remote maintenance system

International Nuclear Information System (INIS)

Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou

1998-01-01

ITER in-vessel components such as blankets are scheduled maintenance components, including complete shield blanket replacement for breeding blankets. In-vessel components are activated by 14 MeV neutrons, so blanket maintenance requires remote handling equipment and tools able to handle heavy payloads of about 4 tons within a positioning accuracy of 2 mm under intense gamma radiation. To facilitate remote maintenance, blankets are segmented into 730 modules and rail-mounted vehicle remote maintenance was developed. According to the ITER R and D program, critical technology related to blanket maintenance was developed extensively through joint efforts of the Japan, EU, and U.S. home teams. This paper summarizes current blanket maintenance technology conducted by the Japan Home Team, including development of full-scale remote handling equipment and tools for blanket maintenance. (author)

Development of blanket remote maintenance system

Energy Technology Data Exchange (ETDEWEB)

Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-04-01

ITER in-vessel components such as blankets are scheduled maintenance components, including complete shield blanket replacement for breeding blankets. In-vessel components are activated by 14 MeV neutrons, so blanket maintenance requires remote handling equipment and tools able to handle heavy payloads of about 4 tons within a positioning accuracy of 2 mm under intense gamma radiation. To facilitate remote maintenance, blankets are segmented into 730 modules and rail-mounted vehicle remote maintenance was developed. According to the ITER R and D program, critical technology related to blanket maintenance was developed extensively through joint efforts of the Japan, EU, and U.S. home teams. This paper summarizes current blanket maintenance technology conducted by the Japan Home Team, including development of full-scale remote handling equipment and tools for blanket maintenance. (author)

Tritium breeding blanket

International Nuclear Information System (INIS)

Smith, D.; Billone, M.; Gohar, Y.; Baker, C.; Mori, S.; Kuroda, T.; Maki, K.; Takatsu, H.; Yoshida, H.; Raffray, A.; Sviatoslavsky, I.; Simbolotti, G.; Shatalov, G.

1991-01-01

The terms of reference for ITER provide for incorporation of a tritium breeding blanket with a breeding ratio as close to unity as practical. A breeding blanket is required to assure an adequate supply of tritium to meet the program objectives. Based on specified design criteria, a ceramic breeder concept with water coolant and an austenitic steel structure has been selected as the first option and lithium-lead blanket concept has been chosen as an alternate option. The first wall, blanket, and shield are integrated into a single unit with separate cooling systems. The design makes extensive use of beryllium to enhance the tritium breeding ratio. The design goals with a tritium breeding ratio of 0.8--0.9 have been achieved and the R ampersand D requirements to qualify the design have been identified. 4 refs., 8 figs., 2 tabs

Fission multipliers for D-D/D-T neutron generators

International Nuclear Information System (INIS)

Lou, T.P.; Vujic, J.L.; Koivunoro, H.; Reijonen, J.; Leung, K.-N.

2003-01-01

A compact D-D/D-T fusion based neutron generator is being designed at the Lawrence Berkeley National Laboratory to have a potential yield of 10 12 D-D n/s and 10 14 D-T n/s. Because of its high neutron yield and compact size (∼20 cm in diameter by 4 cm long), this neutron generator design will be suitable for many applications. However, some applications required higher flux available from nuclear reactors and spallation neutron sources operated with GeV proton beams. In this study, a subcritical fission multiplier with k eff of 0.98 is coupled with the compact neutron generators in order to increase the neutron flux output. We have chosen two applications to show the gain in flux due to the use of fission multipliers--in-core irradiation and out-of-core irradiation. For the in-core irradiation, we have shown that a gain of ∼25 can be achieved in a positron production system using D-T generator. For the out-of-core irradiation, a gain of ∼17 times is obtained in Boron Neutron Capture Therapy (BNCT) using a D-D neutron generator. The total number of fission neutrons generated by a source neutron in a fission multiplier with k eff is ∼50. For the out-of-core irradiation, the theoretical maximum net multiplication is ∼30 due to the absorption of neutrons in the fuel. A discussion of the achievable multiplication and the theoretical multiplication will be presented in this paper

A small scale accelerator driven subcritical assembly development and demonstration experiment at LAMPF

International Nuclear Information System (INIS)

Wender, S.A.; Venneri, F.; Bowman, C.D.; Arthur, E.D.; Heighway, E.A.; Beard, C.A.; Bracht, R.R.; Buksa, J.J.; Chavez, W.; DeVolder, B.G.

1994-01-01

A small scale experiment is described that will demonstrate many of the aspects of accelerator-driven transmutation technology. This experiment uses the high-power proton beam from the Los Alamos Meson Physics Facility accelerator and will be located in the Area-A experimental hall. Beam currents of up to 1 mA will be used to produce neutrons with a molten lead target. The target is surrounded by a molten salt and graphite moderator blanket. Fissionable material can be added to the molten salt to demonstrate plutonium burning or transmutation of commercial spent fuel or energy production from thorium. The experiment will be operated at power levels up to 5 MW t

Neutron measurements in the core and blankets of the reactor Rapsodie; Mesures neutroniques dans le coeur et les couvertures de Rapsodie

Energy Technology Data Exchange (ETDEWEB)

Gourdon, J; Edeline, J C [Commissariat a l' Energie Atomique, 13 - Cadarache (France). Centre d' Etudes Nucleaires

1968-07-01

Beside a brief general discussion, the report contains all the core and blanket neutronic measurements. It covers successively the methods, the measurements themselves and the results. The later concern: spectral indexes, axial and radial fission rates, activation foil measurements and neutronic power determination. (authors) [French] Apres une breve description generale de RAPSODIE, le rapport presente l'ensemble des mesures neutroniques faites dans le coeur et les couvertures. Il traite dans l'ordre des methodes, des mesures et enfin des resultats qui concernent: les indices de spectres, les taux de fission axiaux et radiaux, les mesures par detecteurs par activation, la determination de la puissance, neutronique. (auteurs)

Ferritic steels for the first generation of breeder blankets

International Nuclear Information System (INIS)

Diegele, E.

2009-01-01

Materials development in nuclear fusion for in-vessel components, i.e. for breeder blankets and divertors, has a history of more than two decades. It is the specific in-service and loading conditions and the consequentially required properties in combination with safety standards and social-economic demands that create a unique set of specifications. Objectives of Fusion for Energy (F4E) include: 1) To provide Europe's contribution to the ITER international fusion energy project; 2) To implement the Broader Approach agreement between Euratom and Japan; 3) To prepare for the construction and demonstration of fusion reactors (DEMO). Consequently, activities in F4E focus on structural materials for the first generations of breeder blankets, i.e. ITER Test Blanket Modules (TBM) and DEMO, whereas a Fusion Materials Topical Group implemented under EFDA coordinates R and D on physically based modelling of irradiation effects and R and D in the longer term (new and /or higher risk materials). The paper focuses on martensitic-ferritic steels and (i) reviews briefly the challenges and the rationales for the decisions taken in the past, (ii) analyses the status of the main activities of development and qualification, (iii) indicates unresolved issues, and (iv) outlines future strategies and needs and their implications. Due to the exposure to intense high energy neutron flux, the main issue for breeder materials is high radiation resistance. The First Wall of a breeder blanket should survive 3-5 full power years or, respectively in terms of irradiation damage, typically 50-70 dpa for DEMO and double figures for a power plant. Even though the objective is to have the materials and key fabrication technologies needed for DEMO fully developed and qualified within the next two decades, a major part of the task has to be completed much earlier. Tritium breeding test blanket modules will be installed in ITER with the objective to test DEMO relevant technologies in fusion

Analysis of fission ratio distribution in spherical lithium metal assembly with a graphite reflector

International Nuclear Information System (INIS)

Seki, Yasushi; Maekawa, Hiroshi; Hiraoka, Toru; Hirota, Jitsuya; Moriyama, Masatoshi.

1975-08-01

For the investigation of neutronics in the fusion reactor blanket, a spherical lithium assembly with a graphite reflector was prepared by piling up lithium and graphite blocks. The ratio of U-238 to U-235 fission rates was measured by micro fission chambers, and the result of the measurement was compared with that of the preliminary calculation. It has been shown that there is a large discrepancy between experiment and calculation, which is too large to be explained only by the experimental error. In this paper, the calculational procedure is reviewed and probable causes of the calculational error are listed. Further calculation of the fission ratio is carried out employing refined methods and with varied calculational models. As the result, it is concluded that the cause of the discrepancy is most likely the uncertainties of the nuclear data of the constituent elements of the assembly. (auth.)

High flux transmutation of fission products and actinides

International Nuclear Information System (INIS)

Gerasimov, A.; Kiselev, G.; Myrtsymova, L.

2001-01-01

Long-lived fission products and minor actinides accumulated in spent nuclear fuel of power reactors comprise the major part of high level radwaste. Their incineration is important from the point of view of radwaste management. Transmutation of these nuclides by means of neutron irradiation can be performed either in conventional nuclear reactors, or in specialized transmutation reactors, or in ADS facilities with subcritical reactor and neutron source with application of proton accelerator. Different types of transmutation nuclear facilities can be used in order to insure optimal incineration conditions for radwaste. The choice of facility type for optimal transmutation should be based on the fundamental data in the physics of nuclide transformations. Transmutation of minor actinides leads to the increase of radiotoxicity during irradiation. It takes significant time compared to the lifetime of reactor facility to achieve equilibrium without effective transmutation. High flux nuclear facilities allow to minimize these draw-backs of conventional facilities with both thermal and fast neutron spectrum. They provide fast approach to equilibrium and low level of equilibrium mass and radiotoxicity of transmuted actinides. High flux facilities are advantageous also for transmutation of long-lived fission products as they provide short incineration time

Novel blanket design for ICTR's

International Nuclear Information System (INIS)

Abdel-Khalik, S.I.; Conn, R.W.; Wolfer, W.G.; Larsen, E.N.; Sviatoslavsky, I.N.

1978-01-01

A novel blanket design for ICTRs is described. This blanket is used in SOLASE, the conceptual laser fusion reactor of the University of Wisconsin. The blanket to be described offers numerous advantages, including low cost, low weight, low induced radioactivity levels, the potential for hands-on maintenance, modular construction, low pressure, ability to decouple first wall and blanket coolant temperatures, adequate breeding, low tritium inventory and leakage, and sufficiently long life

Resonance self-shielding effect analysis of neutron data libraries applied for the dual-cooled waste transmutation blanket of the fusion-driven subcritical system

International Nuclear Information System (INIS)

Liu Haibo; Wu Yican; Zheng Shanliang; Zhang Chunzao

2004-01-01

Based on the Fusion-Driven Subcritical System (FDS-I), the 25 groups, 175 groups and 620 groups neutron nuclear data libraries with/without resonance self-shielding correction are made with the Njoy and Transx codes, and the K eff and reaction rates are calculated with the Anisn code. The conclusion indicates that the resonance self-shielding effect affects the reaction rates strongly. (authors)

Mirror reactor blankets

International Nuclear Information System (INIS)

Lee, J.D.; Barmore, W.L.; Bender, D.J.; Doggett, J.N.; Galloway, T.R.

1976-01-01

The general requirements of a breeding blanket for a mirror reactor are described. The following areas are discussed: (1) facility layout and blanket maintenance, (2) heat transfer and thermal conversion system, (3) materials, (4) tritium containment and removal, and (5) nuclear performance

Study of the kinetics parameters for subcritical media driven by source

International Nuclear Information System (INIS)

Lee, S.M.; Maiorino, J.R.

2009-01-01

This paper presents a theoretical study of reactor kinetics focusing on the methodology of calculation and the experimental measurements of the so-called kinetic parameters. A comparison between the methodology based on the Dulla's formalism and the classical method is made. The objective is to exhibit the dependence of the parameters on sub criticality level and perturbation. Two different slab type systems were considered: thermal one and fast one, both with homogeneous media. One group diffusion model was used for the fast reactor, and for the thermal system, two group diffusion model, considering, in both case only one precursor's family. For reason of simplicity, several additional assumptions were made for calculation of two group method: no up-scattering, fission reaction occurring only in thermal group, etc. The solutions for subcritical systems were obtained using the expansion method, and for critical systems, the methods presented in classical textbooks of reactor physics were applied. The numerical results presented their dependence on sub criticality level and perturbation. (author)

Safety features of subcritical fluid fueled systems

International Nuclear Information System (INIS)

Bell, C.R.

1995-01-01

Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible

Safety features of subcritical fluid fueled systems

International Nuclear Information System (INIS)

Bell, C.R.

1994-01-01

Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved in very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible

Safety features of subcritical fluid fueled systems

Energy Technology Data Exchange (ETDEWEB)

Bell, C.R. [Los Alamos National Laboratory, NM (United States)

1995-10-01

Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.

An evaluation of fast reactor blankets

International Nuclear Information System (INIS)

Oosterkamp, W.J.

1974-01-01

A comparative study of different types of fast reactor radial blankets is presented. Included are blankets of fertile material UO 2 , THO 2 and Th-metal blankets of pure reflectors C, BeO, Ni and combinations of reflecting and fertile blankets. The results for 1000MWe cores indicate that there is no incentive to use other than fertile blankets. The most favorable fertile material is thorium due to the prospective higher price of U-233

Comparative analysis of sub-critical transmutation reactor concepts

International Nuclear Information System (INIS)

Chang, S. H.

1997-01-01

The long-lived nuclear wastes have been substantially generated from the light water reactor for a few decades. The toxicity of these spent fuels will be higher than that of the uranium ore, even if those will be stored in the repository more than ten thousands. Hence the means of transmuting the key long-lived nuclear wastes, primarily the minor actinides, using a hybrid proton accelerator and subcritical transmutation reactor, are proposed. Until now, the representative concepts for a subcritical transmutation reactor are the Energy Amplifier, the OMEGA project, the ATW and the MSBR. The detailed concepts and the specifications are illustrated in Table 1. The design requirements for the subcritical transmutation reactor are the high transmutation rate of long-lived nuclear wastes, safety and economics. And to propose the subcritical transmutation reactor concepts, the coolant, the target material and fuel type are carefully considered. In these aspects, the representative concepts for a subcritical transmutation reactor in Table 1 have been surveyed. The requirements for a target and a coolant are the reliable, low maintenance operation and safe operation to minimize the wastes. The reliable, low maintenance operation and safe operation to minimize the wastes. The reliable coolant must have the low melting point, high heat capacity and excellent physical properties. And the target material must have high neutron yield for a given proton condition and easy heat removal capability. Therefore in respect with the above requirements, Pb-Bi is proposed as the coolant and the target material for the subcritical reactor. Because the neutron yield for a given proton energy increases linearly with mass number up to bismuth but in heavier elements spallation events sharply increase both the neutron and heat outputs, Pb-Bi meets not only such the requirements as the above for the coolant but also those for the coolant and target, the simplification of system can be achieved

Physics analyses of an accelerator-driven sub-critical assembly

Science.gov (United States)

Naberezhnev, Dmitry G.; Gohar, Yousry; Bailey, James; Belch, Henry

2006-06-01

Physics analyses have been performed for an accelerator-driven sub-critical assembly as a part of the Argonne National Laboratory activity in preparation for a joint conceptual design with the Kharkov Institute of Physics and Technology (KIPT) of Ukraine. KIPT has a plan to construct an accelerator-driven sub-critical assembly targeted towards the medical isotope production and the support of the Ukraine nuclear industry. The external neutron source is produced either through photonuclear reactions in tungsten or uranium targets, or deuteron reactions in a beryllium target. KIPT intends using the high-enriched uranium (HEU) for the fuel of the sub-critical assembly. The main objective of this paper is to study the possibility of utilizing low-enriched uranium (LEU) fuel instead of HEU fuel without penalizing the sub-critical assembly performance, in particular the neutron flux level. In the course of this activity, several studies have been carried out to investigate the main choices for the system's parameters. The external neutron source has been characterized and a pre-conceptual target design has been developed. Several sub-critical configurations with different fuel enrichments and densities have been considered. Based on our analysis, it was shown that the performance of the LEU fuel is comparable with that of the HEU fuel. The LEU fuel sub-critical assembly with 200-MeV electron energy and 100-kW electron beam power has an average total flux of ˜2.50×10 13 n/s cm 2 in the irradiation channels. The corresponding total facility power is ˜204 kW divided into 91 and 113 kW deposited in the target and sub-critical assemblies, respectively.

Dual coolant blanket concept

International Nuclear Information System (INIS)

Malang, S.; Schleisiek, K.

1994-11-01

A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

Status of fusion reactor blanket design

International Nuclear Information System (INIS)

Smith, D.L.; Sze, D.K.

1986-02-01

The recent Blanket Comparison and Selection Study (BCSS), which was a comprehensive evaluation of fusion reactor blanket design and the status of blanket technology, serves as an excellent basis for further development of blanket technology. This study provided an evaluation of over 130 blanket concepts for the reference case of electric power producing, DT fueled reactors in both Tokamak and Tandem Mirror (TMR) configurations. Based on a specific set of reactor operating parameters, the current understanding of materials and blanket technology, and a uniform evaluation methodology developed as part of the study, a limited number of concepts were identified that offer the greatest potential for making fusion an attractive energy source

Evaluation of Importance of Source Neutrons in Accelerator-Driven System

International Nuclear Information System (INIS)

Kim, Yong Hee; Park, Won Seok

2002-01-01

An importance function of the external spallation neutrons in ADS (Accelerator-Driven System) is defined to characterize the source multiplication in subcritical blanket. For a model ADS problem, the source importance function is evaluated with the TRANSX/TWODANT code system. In order to assess the impact of the power distribution on the importance function, both homogeneous and heterogeneous cores are analyzed and corresponding source multiplications are compared. Also, based on the source importance function, an optimization of the shape of the proton current is performed from the source multiplication point of view. Additionally, the source importance function is compared with the conventional λ-mode adjoint flux, which is used as an importance function of fission neutrons in the critical reactors. Concerning an issue in the ADS design, i.e., difficulty in reducing the fission power unless the proton current is shut off, a study is performed to minimize the source importance, thereby minimizing the fission power, even when the k-eff value of the core is quite high. (authors)

Thermal Hydraulic Analysis of K-DEMO Single Blanket Module for Preliminary Accident Analysis using MELCOR

Energy Technology Data Exchange (ETDEWEB)

Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

2016-05-15

To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

Axisymmetric Magnetic Mirror Fusion-Fission Hybrid

Energy Technology Data Exchange (ETDEWEB)

Moir, R. W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Martovetsky, N. N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Molvik, A. W. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Ryutov, D. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Simonen, T. C. [Univ. of California, Berkeley, CA (United States)

2011-05-13

sale of both electricity and production of fissile fuel with fissioning blankets might need Q~1 while suppressing fissioning might be the most economical application of fusion but will require Q≥4.

Candidate molten salt investigation for an accelerator driven subcritical core

Science.gov (United States)

Sooby, E.; Baty, A.; Beneš, O.; McIntyre, P.; Pogue, N.; Salanne, M.; Sattarov, A.

2013-09-01

We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated. A special thanks is due to Prof. Paul Madden for introducing the ADSMS group to the concept of using the molten salt as the spallation target, rather than a conventional heavy metal spallation target. This feature helps to optimize this core as a Pu/TRU burner.

Subcriticality determination of nuclear fuel assembly by Mihalczo method

International Nuclear Information System (INIS)

Yamane, Yoshihiro; Watanabe, Shoji; Nishina, Kojiro; Miyoshi, Yoshinori; Suzaki, Takenori; Kobayashi, Iwao.

1986-01-01

To establish a technique of on-site subcriticality determination suitable for the criticality safety management of nuclear fuel assembly, the applicability of the method proposed by Mihalczo was examined with the Tank-type Critical Assembly (TCA) of the Japan Atomic Energy Research Institute. In the Mihalczo method, cross power spectral densities and auto power spectral densities are evaluated from the output currents of an ionization chamber containing 252 Cf neutron source and two neutron detectors. The principle of this method is that the spectral ratio formed by the power spectral densities mentioned can be related to the subcriticality by the help of a stochastic theory. Throughout our data processing, an improved formula taking account of the neutron extinction at a detection process was used. Up to the subcriticality of 15 dollars, the Mihalczo method agreed with the water-level worth method, which has been a standard method of reactivity determination at the TCA facility. The systems treated in the present report hold symmetry concerning the nuclear fuel configuration and the 252 Cf chamber position. It was clarified that, contrary to Mihalczo's assertion, the factor converting the spectral ratio to a subcriticality depends on subcriticality itself. (author)

Continuous reactivity calculation for subcritical system

International Nuclear Information System (INIS)

Silva, Cristiano; Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C. da

2011-01-01

With the rise of a new generation of nuclear reactors as for existence the ADS (Accelerator-Driven System), it is important to have a fast and accurate prediction of the variation in reactivity during a possible variation in the intensity of external sources. This paper presents a formulation for the calculation of reactivity in subcritical systems using the inverse method related only to nuclear power derivatives. One of the applications of the proposed method is the possibility of developing reactimeters that allow the continuous monitoring of subcritical systems. (author)

Continuous reactivity calculation for subcritical system

Energy Technology Data Exchange (ETDEWEB)

Silva, Cristiano; Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C. da, E-mail: cristiano@herzeleid.net, E-mail: aquilino@lmp.ufrj.br, E-mail: fernando@con.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2011-07-01

With the rise of a new generation of nuclear reactors as for existence the ADS (Accelerator-Driven System), it is important to have a fast and accurate prediction of the variation in reactivity during a possible variation in the intensity of external sources. This paper presents a formulation for the calculation of reactivity in subcritical systems using the inverse method related only to nuclear power derivatives. One of the applications of the proposed method is the possibility of developing reactimeters that allow the continuous monitoring of subcritical systems. (author)

Reactor Dynamics Experiments with a Sub-Critical Assembly

International Nuclear Information System (INIS)

Miley, G.H.; Yang, Y.; Wu, L.; Momota, H.

2004-01-01

A resurgence in use of nuclear power is now underway worldwide. However due to the shutdown of many university research reactors , student laboratories must rely more heavily on use of sub-critical assemblies. Here a driven sub-critical is described that uses a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The small IEC neutron source would be inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory

CSER-00-007 Addendum 1 Criticality Safety Evaluation of Shippingport PWR Core 2 Blanket Fuel Assemblies at Lower Exposures

International Nuclear Information System (INIS)

WITTEKIND, W.D.

2001-01-01

This analysis meets the requirements of HNF-7098, Criticality Safety Program, (FH 2001a). HNF-7098 states that before starting a new operation with fissile material or before an existing operation is changed, it shall be determined that the entire process will be subcritical under both normal and credible abnormal conditions. To demonstrate the Incredibility Principle is satisfied, this Criticality Safety Evaluation Report (CSER) shows that the form or distribution is such that criticality is impossible. This evaluation demonstrated, that on the basis of effective 235 U enrichment, criticality is not possible. The minimum blanket assembly exposure is 4,375 MW t d/MTU for fissile material that is shown to fulfill the Incredibility Principle safety criterion on the basis of enrichment

Mechanical weathering and rock erosion by climate-dependent subcritical cracking

Science.gov (United States)

Eppes, Martha-Cary; Keanini, Russell

2017-06-01

This work constructs a fracture mechanics framework for conceptualizing mechanical rock breakdown and consequent regolith production and erosion on the surface of Earth and other terrestrial bodies. Here our analysis of fracture mechanics literature explicitly establishes for the first time that all mechanical weathering in most rock types likely progresses by climate-dependent subcritical cracking under virtually all Earth surface and near-surface environmental conditions. We substantiate and quantify this finding through development of physically based subcritical cracking and rock erosion models founded in well-vetted fracture mechanics and mechanical weathering, theory, and observation. The models show that subcritical cracking can culminate in significant rock fracture and erosion under commonly experienced environmental stress magnitudes that are significantly lower than rock critical strength. Our calculations also indicate that climate strongly influences subcritical cracking—and thus rock weathering rates—irrespective of the source of the stress (e.g., freezing, thermal cycling, and unloading). The climate dependence of subcritical cracking rates is due to the chemophysical processes acting to break bonds at crack tips experiencing these low stresses. We find that for any stress or combination of stresses lower than a rock's critical strength, linear increases in humidity lead to exponential acceleration of subcritical cracking and associated rock erosion. Our modeling also shows that these rates are sensitive to numerous other environment, rock, and mineral properties that are currently not well characterized. We propose that confining pressure from overlying soil or rock may serve to suppress subcritical cracking in near-surface environments. These results are applicable to all weathering processes.

(D,T) Driven thorium hybrid blankets

International Nuclear Information System (INIS)

Al-Kusayer, T.A.; Khan, S.; Sahin, S.

1983-01-01

Recently, a project has started, with the aim to establish the neutronic performance and the basic design of an experimental fusionfission (hybrid) reactor facility, called AYMAN, in cylinderical geometry. The fusion reactor will have to be simulated by a (D,T) neutron generator. Fissile and fertile fuel will have to surround the neutron generator as a cylinderical blanket to simulate the boundary conditions of the hybrid blanket in a proper way. This geometry is consistent with Tandem Mirror Hybrid Blanket design and with most of the ICF blanket designs. A similar experimental installation will become operational around 1984 at the Swiss Federal Institute of Technology in Lausanne, Switzerland known under the project LOTUS. Due to the limited dimensions of the experimental cavity of the LOTUS-hybrid reactor, the LOTUS blankets have to be designed in plane geometry. Also, the bulky form of the Haefely neutron generator of the LOTUS facility obliges one to design a blanket in the plane geometry. This results in a vacuum left boundary conditions for the LOTUS blanket. The importance of a reflecting left boundary condition on the overall neutronic performance of a hybrid blanket has been analyzed in previous work in detail

Concepts for fusion fuel production blankets

International Nuclear Information System (INIS)

Gierszewski, P.

1986-06-01

The fusion blanket surrounds the burning hydrogen core of the fusion reactor. It is in this blanket that most of the energy released by the DT fusion reaction is converted into useable product, and where tritium fuel is produced to enable further operation of the reactor. Blankets will involve new materials, conditions and processes. Several recent fusion blanket concepts are presented to illustrate the range of ideas

A blanket design, apparatus, and fabrication techniques for the mass production of multilayer insulation blankets for the Superconducting Super Collider

International Nuclear Information System (INIS)

Gonczy, J.D.; Boroski, W.N.; Niemann, R.C.; Otavka, J.G.; Ruschman, M.K.; Schoo, C.J.

1989-09-01

The multilayer insulation (MLI) system for the Superconducting Super Collider (SSC) consists of full cryostat length assemblies of aluminized polyester film fabricated in the form of blankets and installed as blankets to the 4.5K cold mass and the 20K and 80K thermal radiation shields. Approximately 40,000 MLI blankets will be required in the 10,000 cryogenic devices comprising the SSC accelerator. Each blanket is nearly 17 meters long and 1.8 meters wide. This paper reports the blanket design, an apparatus, and the fabrication method used to mass produce pre-fabricated MLI blankets. Incorporated in the blanket design are techniques which automate quality control during installation of the MLI blankets in the SSC cryostat. The apparatus and blanket fabrication method insure consistency in the mass produced blankets by providing positive control of the dimensional parameters which contribute to the thermal performance of the MLI blanket. By virtue of the fabrication process, the MLI blankets have inherent features of dimensional stability three-dimensional uniformity, controlled layer density, layer-to-layer registration, interlayer cleanliness, and interlayer material to accommodate thermal contraction differences. 11 refs., 6 figs., 1 tab

Breeding blanket for Demo

International Nuclear Information System (INIS)

Proust, E.; Giancarli, L.

1992-01-01

This paper presents the main design features, their rationale, and the main critical issues for the development, of the four DEMO-relevant blanket concepts presently investigated within the framework of the European Test-Blanket Development Programme

Subcriticality of accelerator driven system by AESJ/JAERI working party

International Nuclear Information System (INIS)

Iwasaki, Tomohiko

2002-01-01

Under Atomic Energy Society of Japan (AESJ) and Japan Atomic Energy Research Institute (JAERI), a Working Party on Reactor Physics of Accelerator-Driven System (ADS-WP) has been set since March 1999 to review and investigate special subjects related to reactor physics research of Accelerator-Driven System (ADS). In the ADS-WP, the extensive and aggressive activity is being made by 25 professional members in the field of reactor physics in Japan. The ADS is now studying three subjects related to subcriticality of ADS; (1) calculation accuracy of subcriticality on ADS, (2) critical safety issues of ADS, and (3) theoretical review of subcriticality and its measurement methods. This paper describes two topics related to the subjects (1) and (2); one is an analysis of maximum reactivity potentially inserted to a subcritical core and the other is a benchmark proposal for checking calculation accuracy of subcriticality on ADS. The full specification of the calculation benchmark will be supplied by June 2002. Researchers from overseas, especially from Korea, are welcome to join this benchmark

Neutron chain length distributions in subcritical systems

International Nuclear Information System (INIS)

Nolen, S.D.; Spriggs, G.

1999-01-01

In this paper, the authors present the results of the chain-length distribution as a function of k in subcritical systems. These results were obtained from a point Monte Carlo code and a three-dimensional Monte Carlo code, MC++. Based on these results, they then attempt to explain why several of the common neutron noise techniques, such as the Rossi-α and Feynman's variance-to-mean techniques, are difficult to perform in highly subcritical systems using low-efficiency detectors

Blankets for thermonuclear device

International Nuclear Information System (INIS)

Maki, Koichi; Fukumoto, Hideshi.

1986-01-01

Purpose: To produce tritium more than consumed, through thermonuclear reaction. Constitution: The energy spectrum of neutron generated by neutron multiplying reaction in a neutron multiplying blanket and moderated neutrons has a large ratio in a low energy section. In the low-energy absorption region of stainless steel which is a material of cooling pipes constituting a neutron multiplying blanket cooling channel, the neutrons are absorbed, lessening the neutron multiplying effect. To prevent this, the neutron multiplying blanket cooling channel is covered with tritium breeding blankets, thereby enabling the production of a substantially great amount of tritium more than the amount of tritium to be consumed by the thermonuclear reaction by preventing neutron absorption by the component materials of the cooling channel, improving the tritium breeding ratio by 20 to 25 %, and increasing the efficiency of use of neutrons for tritium generation. (Horiuchi, T.)

Ensuring the validity of calculated subcritical limits

International Nuclear Information System (INIS)

Clark, H.K.

1977-01-01

The care taken at the Savannah River Laboratory and Plant to ensure the validity of calculated subcritical limits is described. Close attention is given to ANSI N16.1-1975, ''Validation of Calculational Methods for Nuclear Criticality Safety.'' The computer codes used for criticality safety computations, which are listed and are briefly described, have been placed in the SRL JOSHUA system to facilitate calculation and to reduce input errors. A driver module, KOKO, simplifies and standardizes input and links the codes together in various ways. For any criticality safety evaluation, correlations of the calculational methods are made with experiment to establish bias. Occasionally subcritical experiments are performed expressly to provide benchmarks. Calculated subcritical limits contain an adequate but not excessive margin to allow for uncertainty in the bias. The final step in any criticality safety evaluation is the writing of a report describing the calculations and justifying the margin

Subcritical Water Hydrolysis of Peptides: Amino Acid Side-Chain Modifications

Science.gov (United States)

Powell, Thomas; Bowra, Steve; Cooper, Helen J.

2017-09-01

Previously we have shown that subcritical water may be used as an alternative to enzymatic digestion in the proteolysis of proteins for bottom-up proteomics. Subcritical water hydrolysis of proteins was shown to result in protein sequence coverages greater than or equal to that obtained following digestion with trypsin; however, the percentage of peptide spectral matches for the samples treated with trypsin were consistently greater than for those treated with subcritical water. This observation suggests that in addition to cleavage of the peptide bond, subcritical water treatment results in other hydrolysis products, possibly due to modifications of amino acid side chains. Here, a model peptide comprising all common amino acid residues (VQSIKCADFLHYMENPTWGR) and two further model peptides (VCFQYMDRGDR and VQSIKADFLHYENPTWGR) were treated with subcritical water with the aim of probing any induced amino acid side-chain modifications. The hydrolysis products were analyzed by direct infusion electrospray tandem mass spectrometry, either collision-induced dissociation or electron transfer dissociation, and liquid chromatography collision-induced dissociation tandem mass spectrometry. The results show preferential oxidation of cysteine to sulfinic and sulfonic acid, and oxidation of methionine. In the absence of cysteine and methionine, oxidation of tryptophan was observed. In addition, water loss from aspartic acid and C-terminal amidation were observed in harsher subcritical water conditions. [Figure not available: see fulltext.

Minimum thickness blanket-shield for fusion reactors

International Nuclear Information System (INIS)

Karni, Y.; Greenspan, E.

1989-01-01

A lower bound on the minimum thickness fusion reactor blankets can be designed to have, if they are to breed 1.267 tritons per fusion neutron, is identified by performing a systematic nucleonic optimization of over a dozen different blanket concepts which use either Be, Li 17 Pb 83 , W or Zr for neutron multiplication. It is found that Be offers minimum thickness blankets; that the blanket and shield (B/S) thickness of Li 17 Pb 83 based blankets which are supplemented by Li 2 O and/or TiH 2 are comparable to the thickness of Be based B/S; that of the Be based blankets, the aqueous self-cooled one offers one of the most compact B/S; and that a number of blanket concepts might enable the design of B/S which is approximately 12 cm and 39 cm thinner than the B/S thickness of, respectively, conventional self-cooled Li 17 Pb 83 and Li blankets. Aqueous self-cooled tungsten blankets could be useful for experimental fusion devices provided they are designed to be heterogeneous. (orig.)

Subcriticality determination in ADS: Valina-Booster experiments

International Nuclear Information System (INIS)

Persson, C. M.; Gudowski, W.; Fokau, A.; Bournos, V.; Fokov, Y.; Routkovskaia, C.; Serafimovich, I.; Kiyavitskaya, H.

2007-01-01

A major problem in operating a full-scale subcritical accelerator-driven system (ADS) is to ensure sufficient margin to criticality. Therefore, reliable techniques for subcriticality monitoring are required. In order to develop such techniques, a full understanding of existing reactivity determination methods is essential. In this work, reactivity determination methods, such as pulsed neutron source methods and noise methods, are studied experimentally in the subcritical facility YALINA-Booster. YALINA-Booster: The subcritical assembly YALINA-Booster: recently constructed at the Joint Institute for Power and Nuclear Research - Sosny, consists of a subcritical core driven by an external neutron source. The neutron source is a powerful neutron generator consisting of a deuteron accelerator and a target of deuterium or tritium embedded in titanium. Through (d, d) - or (d, t)-reactions neutrons are created with energy around 2.5 MeV and 14.1 MeV respectively. Neutrons are born in the centre of the core and multiply through a lead matrix fuelled with highly enriched uranium (90% and 36%). This zone is referred to as the booster zone and is surrounded by a thermal zone, moderated by polyethylene. In order to reach sufficient high effective multiplication factor, the thermal zone is fuelled by approximately one thousand rods of 10% enriched uranium dioxide in cylindrical geometry. To prevent thermal neutrons from diffusing into the fast booster zone, an interface, consisting of boron carbide and natural uranium rods, is located between the zones. YALINA-Booster has a radial graphite reflector of thickness 24 cm. Experiments: Experiments using the neutron source in pulsed mode will be presented, relying on methods such as the area method and the method of prompt neutron decay rate determination. Moreover, results from noise analysis using for instance the Feynman-α method will be presented

Disruption problematics in segmented blanket concepts

International Nuclear Information System (INIS)

Crutzen, Y.; Fantechi, S.; Farfaletti-Casali, F.

1994-01-01

In Tokamaks, the hostile operating environment originated by plasma disruption events requires that the first wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence there is a need to improve the safety features of the blanket design concepts satisfying the disruption problematics and to formulate guidelines on the required internal reinforcements of the blanket components. The present paper describes the recent investigations on blanket reinforcement systems needed in order to optimize the first-wall/blanket/shield structural design for next step and commercial fusion reactors in the context of ITER, DEMO and SEAFP activities

Subcritical tests - nuclear weapon testing under the Comprehensive Test Ban Treaty

International Nuclear Information System (INIS)

Hoeibraaten, S.

1998-10-01

The report discusses possible nuclear weapons related experiments and whether these are permitted under the 1996 Comprehensive Test Ban Treaty (CTBT). The term ''subcritical experiments'' as used in the United States includes experiments in which one studies fissile materials (so far only plutonium) under extreme conditions generated by conventional high explosives, and in which a self-sustained chain reaction never develops in the fissile material. The known facts about the American subcritical experiments are presented. There is very little reason to doubt that these experiments were indeed subcritical and therefore permitted under the CTBT. Little is known about the Russian efforts that are being made on subcritical experiments

The Muse-4 experiment: measurement of the kinetic parameters of a subcritical system; L'experience MUSE-4: mesure des parametres cinetiques d'un systeme sous-critique

Energy Technology Data Exchange (ETDEWEB)

Vollaire, J

2004-10-01

Accelerator Driven Systems (ADS) which are based on an external neutron source coupled to a subcritical core, offer advantages for the incineration of radioactive waste. In order to understand the neutronic specificity of such a system, during the MUSE IV experimental program, the experimental reactor MASURCA (CEA Cadarache) has been coupled to the neutrons source GENEPI. This setup has enabled the development of an on-line measurement technique of the effective multiplication factor. This measurement benefits from the characteristics of the reactor response depending on the multiplication factor at the prompt fission and delayed fission time scales. The analysis of those experiments shows that the proposed method give results in agreement with the one deduced using classical reactivity measurement techniques which can not however be used in a power ADS. (author)

Economic implications of fusion-fission energy systems

International Nuclear Information System (INIS)

Deonigi, D.E.; Schulte, S.C.

1979-04-01

The principal conclusions that can be made based on the estimated costs reported in this paper are twofold. First, hybrid reactors operating symbiotically with conventional fission reactors are a potentially attractive supply alternative. Estimated hybrid energy system costs are slightly greater than estimated costs of the most attractive alternatives. However, given the technological, economic, and institutional uncertainties associated with future energy supply, differences of such magnitude are of little significance. Second, to be economically viable, hybrid reactors must be both fuel producers and electricity producers. A data point representing each hybrid reactor driver-blanket concept is plotted as a function of net electrical production efficiency and annual fuel production. The plots illustrate that the most economically viable reactor concepts are those that produce both fuel and electricity

Fusion breeder: its potential role and prospects

International Nuclear Information System (INIS)

Lee, J.D.

1981-01-01

The fusion breeder is a concept that utilizes 14 MeV neutrons from D + T → n(14.1 MeV) + α(3.5 MeV) fusion reactions to produce more fuel than the tritium (T) needed to sustain the fusion process. This excess fuel production capacity is used to produce fissile material (Pu-239 or U-233) for subsequent use in fission reactors. We are concentrating on a class of blankets we call fission suppressed. The blanket is the region surrounding the fusion plasma in which fusion neutrons interact to produce fuel and heat. The fission-suppressed blanket uses non-fission reactions (mainly (n,2n) or (n,n't)) to generate excess neutrons for the production of net fuel. This is in contrast to the fast fission class of blankets which use (n,fiss) reactions to generate excess neutrons. Fusion reactors with fast fission blankets are commony known as fusion-fission hybrids because they combine fusion and fission in the same device

An economic parametric analysis of the synthetic fuel produced by a fusion-fission complex

International Nuclear Information System (INIS)

Tai, A.S.; Krakowski, R.A.

1980-01-01

A simple analytic model is used to examine economic constraints of a fusion-fission complex in which a portion of a thermal energy is used for producing synthetic fuel (synfuel). Since the values of many quantities are not well-known, a parametric analysis has been carried out for testing the sensitivity of the synfuel production cost in relation to crucial economic and technological quantities (investment costs of hybrid and synfuel plants, energy multiplication of the fission blanket, recirculating power fraction of the fusion driver, etc.). In addition, a minimum synfuel selling price has been evaluated, from which the fission-fusion-synfuel complex brings about a higher economic benefit than does the fusion-fission hybrid entirely devoted to fissile-fuel and electricity generation. This paper describes the energy flow diagram of fusion-fission synfuel concept, express the revenue-to-cost formulation and the breakeven synfuel selling price. The synfuel production cost given by the model is evaluated within a range of values of crucial parameters. Assuming an electric cost of 2.7 cents/kWh, an annual investment cost per energy unit of 4.2 to 6 $/FJ for the fusion-fission complex and 1.5 to 3 $/GJ for the synfuel plant, the synfuel production cost lies between 6.5 and 8.5 $/GJ. These production costs can compete with those evaluated for other processes. The study points out a potential use of the fusion-fission hybrid reactor for other than fissile-fuel and electricity generation. (orig.) [de

ARIES-IV Nested Shell Blanket Design

International Nuclear Information System (INIS)

Wong, C.P.C.; Redler, K.; Reis, E.E.; Will, R.; Cheng, E.; Hasan, C.M.; Sharafat, S.

1993-11-01

The ARIES-IV Nested Shell Blanket (NSB) Design is an alternate blanket concept of the ARIES-IV low activation helium-cooled reactor design. The reference design has the coolant routed in the poloidal direction and the inlet and outlet plena are located at the top and bottom of the torus. The NSB design has the high velocity coolant routed in the toroidal direction and the plena are located behind the blanket. This is of significance since the selected structural material is SiC-composite. The NSB is designed to have key high performance components with characteristic dimensions of no larger than 2 m. These components can be brazed to form the blanket module. For the diverter design, we eliminated the use of W as the divertor coating material by relying on the successful development of the gaseous divertor concept. The neutronics and thermal-hydraulic performance of both blanket concepts are similar. The selected blanket and divertor configurations can also meet all the projected structural, neutronics and thermal-hydraulics design limits and requirements. With the selected blanket and divertor materials, the design has a level of safety assurance rate of I (LSA-1), which indicates an inherently safe design

U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

International Nuclear Information System (INIS)

1995-01-01

Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li 2 O) and lithium zirconate (Li 2 ZrO 3 ) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers

LMFBR blanket physics project progress report No. 4

International Nuclear Information System (INIS)

Driscoll, M.J.; Lanning, D.D.; Kaplan, I.; Supple, A.T.

1973-01-01

During the period covered by the report, July 1, 1972, through June 30, 1973, work was devoted to completion of experimental measurements and data analysis on Blanket Mockup No. 3, a graphite-reflected blanket, and to initiation of experimental work on Blanket Mockup No. 4, a steel-reflected assembly designed to mock up a demonstration plant blanket. Work was also carried out on the analysis of a number of advanced blanket concepts, including the use of high-albedo reflectors, the use of thorium in place of uranium in the blanket region, and the ''parfait'' or completely internal blanket concept. Finally, methods development work was initiated to develop the capability for making gamma heating measurements in the blanket mockups. (U.S.)

Neutronic studies of fissile and fusile breeding blankets

International Nuclear Information System (INIS)

Taczanowski, S.

1984-08-01

In light of the need of convincing motivation substantiating expensive and inherently applied research (nuclear energy), first a simple comparative study of fissile breeding economics of fusion fission hybrids, spallators and also fast breeder reactors has been carried out. As a result, the necessity of maximization of fissile production (in the first two ones, in fast breeders rather the reprocessing costs should be reduced) has been shown, thus indicating the design strategy (high support ratio) for these systems. In spite of the uncertainty of present projections onto further future and discrepancies in available data even quite conservative assumptions indicate that hybrids and perhaps even earlier - spallators can become economic at realistic uranium price increase and successfully compete against fast breeders. Then on the basis of the concept of the neutron flux shaping aimed at the correlation of the selected cross-sections with the neutron flux, the indications for the maximization of respective reaction rates has been formulated. In turn, these considerations serve as the starting point for the guidelines of breeding blanket nuclear design, which are as follows: 1) The source neutrons must face the multiplying layer (of proper thickness) of possibly low concentration of nuclides attenuating the neutron multiplication (i.e. structure materials, nongaseous coolants). 2) For the most effective trapping of neutrons within the breeding zone (leakage and void streaming reduction) it must contain an efficient moderator (not valid for fissile breeding blankets). 3) All regions of significant slow flux should contain 6 Li in order to reduce parasite neutron captures in there. (orig./HP)

Self-cooled liquid-metal blanket concept

International Nuclear Information System (INIS)

Malang, S.; Arheidt, K.; Barleon, L.

1988-01-01

A blanket concept for the Next European Torus (NET) where 83Pb-17Li serves both as breeder material and as coolant is described. The concept is based on the use of novel flow channel inserts for a decisive reduction of the magnetohydrodynamic (MHD) pressure drop and employs beryllium as neutron multiplier in order to avoid the need for breeding blankets at the inboard side of the torus. This study includes the design, neutronics, thermal hydraulics, stresses, MHDs, corrosion, tritium recovery, and safety of a self-cooled liquid-metal blanket. The results of the investigations indicate that the self-cooled blanket is an attractive alternative to other driver blanket concepts for NET and that it can be extrapolated to the conditions of a DEMO reactor

Fission theory and actinide fission data

Energy Technology Data Exchange (ETDEWEB)

Michaudon, A.

1975-06-01

The understanding of the fission process has made great progress recently, as a result of the calculation of fission barriers, using the Strutinsky prescription. Double-humped shapes were obtained for nuclei in the actinide region. Such shapes could explain, in a coherent manner, many different phenomena: fission isomers, structure in near-threshold fission cross sections, intermediate structure in subthreshold fission cross sections and anisotropy in the emission of the fission fragments. A brief review of fission barrier calculations and relevant experimental data is presented. Calculations of fission cross sections, using double-humped barrier shapes and fission channel properties, as obtained from the data discussed previously, are given for some U and Pu isotopes. The fission channel theory of A. Bohr has greatly influenced the study of low-energy fission. However, recent investigation of the yields of prompt neutrons and γ rays emitted in the resonances of {sup 235}U and {sup 239}Pu, together with the spin determination for many resonances of these two nuclei cannot be explained purely in terms of the Bohr theory. Variation in the prompt neutron and γ-ray yields from resonance to resonance does not seem to be due to such fission channels, as was thought previously, but to the effect of the (n,γf) reaction. The number of prompt fission neutrons and the kinetic energy of the fission fragments are affected by the energy balance and damping or viscosity effects in the last stage of the fission process, from saddle point to scission. These effects are discussed for some nuclei, especially for {sup 240}Pu.

Subcritical calculation of the nuclear material warehouse

International Nuclear Information System (INIS)

Garcia M, T.; Mazon R, R.

2009-01-01

In this work the subcritical calculation of the nuclear material warehouse of the Reactor TRIGA Mark III labyrinth in the Mexico Nuclear Center is presented. During the adaptation of the nuclear warehouse (vault I), the fuel was temporarily changed to the warehouse (vault II) and it was also carried out the subcritical calculation for this temporary arrangement. The code used for the calculation of the effective multiplication factor, it was the Monte Carlo N-Particle Extended code known as MCNPX, developed by the National Laboratory of Los Alamos, for the particles transport. (Author)

Blanket comparison and selection study. Volume II

International Nuclear Information System (INIS)

1983-10-01

This volume contains extensive data for the following chapters: (1) solid breeder tritium recovery, (2) solid breeder blanket designs, (3) alternate blanket concept screening, and (4) safety analysis. The following appendices are also included: (1) blanket design guidelines, (2) power conversion systems, (3) helium-cooled, vanadium alloy structure blanket design, (4) high wall loading study, and (5) molten salt safety studies

Fusion blanket inherent safety assessment

International Nuclear Information System (INIS)

Sze, D.K.; Jung, J.; Cheng, E.T.

1986-01-01

Fusion has significant potential safety advantages. There is a strong incentive for designing fusion plants to ensure that inherent safety will be achieved. Accordingly, both the Tokamak Power Systems Studies and MINIMARS have identified inherent safety as a design goal. A necessary condition is for the blanket to maintain its configuration and integrity under all credible accident conditions. A main problem is caused by afterheat removal in an accident condition. In this regard, it is highly desirable to achieve the required level of protection of the plant capital investment and limitation of radioactivity release by systems that rely only on inherent properties of matter (e.g., thermal conductivity, specific heat, etc.) and without the use of active safety equipment. This paper assesses the conditions under which inherent safety is feasible. Three types of accident conditions are evaluated for two blankets. The blankets evaluated are a self cooled vanadium/lithium blanket and a self-cooled vanadium/Flibe blanket. The accident conditions evaluated are: (1) loss-of-flow accident; (2) loss-of-coolant accident (LOCA); and (3) partial loss-of-coolant accident

Blanket materials for DT fusion reactors

International Nuclear Information System (INIS)

Smith, D.L.

1981-01-01

This paper presents an overview of the critical materials issues that must be considered in the development of a tritium breeding blanket for a tokamak fusion reactor that operates on the D-T-Li fuel cycle. The primary requirements of the blanket system are identified and the important criteria that must be considered in the development of blanket technology are summarized. The candidate materials are listed for the different blanket components, e.g., breeder, coolant, structure and neutron multiplier. Three blanket concepts that appear to offer the most potential are: (1) liquid-metal breeder/coolant, (2) liquid-metal breeder/separate coolant, and (3) solid breeder/separate coolant. The major uncertainties associated with each of the design concepts are discussed and the key materials R and D requirements for each concept are identified

Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

Energy Technology Data Exchange (ETDEWEB)

Greenspan, Ehud [Univ. of California, Berkeley, CA (United States)

2015-11-04

This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective of this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and

Blanket maintenance by remote means using the cassette blanket approach

International Nuclear Information System (INIS)

Werner, R.W.

1978-01-01

Induced radioactivity in the blanket and other parts of a fusion reactor close to the plasma zone will dictate remote assembly, disassembly, and maintenance procedures. Time will be of the essence in these procedures. They must be practicable and certain. This paper discusses the reduction of a complicated Tokamak reactor to a simpler assembly via the use of a vacuum building in which to house the reactor and the introduction in this new model of cassette blanket modules. The cassettes significantly simplify remote handling

Reaction kinetics of cellulose hydrolysis in subcritical and supercritical water

Science.gov (United States)

Olanrewaju, Kazeem Bode

The uncertainties in the continuous supply of fossil fuels from the crisis-ridden oil-rich region of the world is fast shifting focus on the need to utilize cellulosic biomass and develop more efficient technologies for its conversion to fuels and chemicals. One such technology is the rapid degradation of cellulose in supercritical water without the need for an enzyme or inorganic catalyst such as acid. This project focused on the study of reaction kinetics of cellulose hydrolysis in subcritical and supercritical water. Cellulose reactions at hydrothermal conditions can proceed via the homogeneous route involving dissolution and hydrolysis or the heterogeneous path of surface hydrolysis. The work is divided into three main parts. First, the detailed kinetic analysis of cellulose reactions in micro- and tubular reactors was conducted. Reaction kinetics models were applied, and kinetics parameters at both subcritical and supercritical conditions were evaluated. The second major task was the evaluation of yields of water soluble hydrolysates obtained from the hydrolysis of cellulose and starch in hydrothermal reactors. Lastly, changes in molecular weight distribution due to hydrothermolytic degradation of cellulose were investigated. These changes were also simulated based on different modes of scission, and the pattern generated from simulation was compared with the distribution pattern from experiments. For a better understanding of the reaction kinetics of cellulose in subcritical and supercritical water, a series of reactions was conducted in the microreactor. Hydrolysis of cellulose was performed at subcritical temperatures ranging from 270 to 340 °C (tau = 0.40--0.88 s). For the dissolution of cellulose, the reaction was conducted at supercritical temperatures ranging from 375 to 395 °C (tau = 0.27--0.44 s). The operating pressure for the reactions at both subcritical and supercritical conditions was 5000 psig. The results show that the rate-limiting step in

U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

Energy Technology Data Exchange (ETDEWEB)

1995-01-01

Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li{sub 2}O) and lithium zirconate (Li{sub 2}ZrO{sub 3}) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers.

Study on uranium-water multiplicative means of the (RESUCO-Subcritical experimental reactor of uranium with oxygen) subcritical assembly by pulsed neutron technique

International Nuclear Information System (INIS)

Jesus Barbosa, S. de.

1987-01-01

The effective multiplication factor and the nuclear parameters associated with the variation of (RESUCO- Subcritical Experimental Reactor of Uranium with Oxygen) Subcritical Assembly Configuration, using pulsed neutron technique are analysed. BF3 detectors were used to detect the variation of thermal neutrons in the system, positioned parallelly to fuel elements, and a proton recoil detector was used for monitoring the neutron generation. (M.C.K.) [pt

Design requirement on HYPER blanket fuel assembly

International Nuclear Information System (INIS)

Hwang, Woan; Lee, B. O.; Nam, C.; Ryu, W. S.; Lee, B. S.; Park, W. S.

2000-07-01

This document describes design requirements which are needed for designing the blanket assembly of the HYPER as design guidance. The blanket assembly of the HYPER consists of blanket fuel rods, mounting rail, spacer, upper nozzle with handling socket, bottom nozzle with mounting rail and skeleton structure. The blanket fuel rod consists of top end plug, bottom end plug with key way, blanket fuel slug, and cladding. In the assembly, the rods are in a triangular pitch array. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements for the blanket fuel assembly of the HYPER

Fusion--fission hybrid reactors based on the laser solenoid

International Nuclear Information System (INIS)

Steinhauer, L.C.; Taussig, R.T.; Quimby, D.C.

1976-01-01

Fusion-fission reactors, based on the laser solenoid concept, can be much smaller in scale than their pure fusion counterparts, with moderate first-wall loading and rapid breeding capabilities (1 to 3 tonnes/yr), and can be designed successfully on the basis of classical plasma transport properties and free-streaming end-loss. Preliminary design information is presented for such systems, including the first wall, pulse coil, blanket, superconductors, laser optics, and power supplies, accounting for the desired reactor performance and other physics and engineering constraints. Self-consistent point designs for first and second generation reactors are discussed which illustrate the reactor size, performance, component parameters, and the level of technological development required

Evaluation of DD and DT fusion fuel cycles for different fusion-fission energy systems

International Nuclear Information System (INIS)

Gohar, Y.

1980-01-01

A study has been carried out in order to investigate the characteristics of an energy system to produce a new source of fissile fuel for existing fission reactors. The denatured fuel cycles were used because it gives additional proliferation resistance compared to other fuel cycles. DT and DD fusion drivers were examined in this study with a thorium or uranium blanket for each fusion driver. Various fuel cycles were studied for light-water and heavy-water reactors. The cost of electricity for each energy system was calculated

Subcritical reactivity measurement at Angra 1 nuclear power plant

International Nuclear Information System (INIS)

Kuramoto, Renato Yoichi Ribeiro; Miranda, Anselmo Ferreira

2011-01-01

In order to speed up the Angra 1 NPP physics tests, this work intends to develop a digital reactivity meter combined with a methodology of the modified Neutron Source Multiplication (NSM) method with correction factors for subcriticality measurements at Angra 1 NPP. In the first part of this work, we have applied the Modified Neutron Source Multiplication (MNSM) Method with fundamental mode extraction, in order to improve the monitoring of the subcriticality at Angra 1 NPP during the criticality approach. In the second part, we developed a preliminary subcritical reactivity meter algorithm based on the point-reactor inverse kinetic model with six delayed neutron groups and external neutron source. The source strength was obtained through the Least Squares Inverse Kinetics Method (LSIKM). (author)

Fusion-reactor blanket and coolant material compatibility

International Nuclear Information System (INIS)

Jeppson, D.W.; Keough, R.F.

1981-01-01

Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

Physics of subcritical multiplying regions and experimental validation

International Nuclear Information System (INIS)

Salvatores, M.

1996-01-01

The coupling of a particle accelerator with a spallation target and with a subcritical multiplying region has been proposed in the fifties and is called here a hybrid system. This article gives some ideas about the energetic balance of such a system. The possibilities of experimental validation of some properties of a subcritical multiplying region by using MASURCA facility at CEA-Cadarache are examined. The results of a preliminary experiment called MUSE are presented. (A.C.)

Pulsed neutron source based on accelerator-subcritical-assembly

Energy Technology Data Exchange (ETDEWEB)

Inoue, Makoto; Noda, Akira; Iwashita, Yoshihisa; Okamoto, Hiromi; Shirai, Toshiyuki [Kyoto Univ., Uji (Japan). Inst. for Chemical Research

1997-03-01

A new pulsed neutron source which consists of a 300MeV proton linac and a nuclear fuel subcritical assembly is proposed. The proton linac produces pulsed spallation neutrons, which are multipied by the subcritical assembly. A prototype proton linac that accelerates protons up to 7MeV has been developed and a high energy section of a DAW structure is studied with a power model. Halo formations in high intensity beam are also being studied. (author)

Spallator and APEX nuclear fuel cycle: a new option for nuclear power

International Nuclear Information System (INIS)

Steinberg, M.

1982-01-01

A new nuclear fuel cycle is described which provides a long term supply of nuclear fuel for the thermal LWR nuclear power reactors and eliminates the need for long-term storage of radioactive waste. Fissile fuel is produced by the Spallator which depends on the production of spallation neutrons by the interaction of high-energy (1 to 2 GeV) protons on a heavy-metal target. The neutrons are absorbed in a surrounding natural-uranium or thorium blanket in which fissile Pu-239 to U-233 is produced. Advances in linear accelerator technology makes it possible to design and construct a high-beam-current continuous-wave proton linac for production purposes. The target is similar to a sub-critical reactor and produces heat which is converted to electricity for supplying the linac. The Spallator is a self-sufficient fuel producer, which can compete with the fast breeder. The APEX fuel cycle depends on recycling the transuranics and long-lived fission products while extracting the stable and short-lived fission products when reprocessing the fuel. Transmutation and decay within the fuel cycle and decay of short-lived fission products external to the fuel cycle eliminates the need for long-term geological age shortage of fission-product waste

Spallator and APEX nuclear fuel cycle: a new option for nuclear power

Energy Technology Data Exchange (ETDEWEB)

Steinberg, M.

1982-01-01

A new nuclear fuel cycle is described which provides a long term supply of nuclear fuel for the thermal LWR nuclear power reactors and eliminates the need for long-term storage of radioactive waste. Fissile fuel is produced by the Spallator which depends on the production of spallation neutrons by the interaction of high-energy (1 to 2 GeV) protons on a heavy-metal target. The neutrons are absorbed in a surrounding natural-uranium or thorium blanket in which fissile Pu-239 to U-233 is produced. Advances in linear accelerator technology makes it possible to design and construct a high-beam-current continuous-wave proton linac for production purposes. The target is similar to a sub-critical reactor and produces heat which is converted to electricity for supplying the linac. The Spallator is a self-sufficient fuel producer, which can compete with the fast breeder. The APEX fuel cycle depends on recycling the transuranics and long-lived fission products while extracting the stable and short-lived fission products when reprocessing the fuel. Transmutation and decay within the fuel cycle and decay of short-lived fission products external to the fuel cycle eliminates the need for long-term geological age shortage of fission-product waste.

Low technology high tritium breeding blanket concept

International Nuclear Information System (INIS)

Gohar, Y.; Baker, C.C.; Smith, D.L.

1987-10-01

The main function of this low technology blanket is to produce the necessary tritium for INTOR operation with minimum first wall coverage. The INTOR first wall, blanket, and shield are constrained by the dimensions of the reference design and the protection criteria required for different reactor components and dose equivalent after shutdown in the reactor hall. It is assumed that the blanket operation at commercial power reactor conditions and the proper temperature for power generation can be sacrificed to achieve the highest possible tritium breeding ratio with minimum additional research and developments and minimal impact on reactor design and operation. A set of blanket evaluation criteria has been used to compare possible blanket concepts. Six areas: performance, operating requirements, impact on reactor design and operation, safety and environmental impact, technology assessment, and cost have been defined for the evaluation process. A water-cooled blanket was developed to operate with a low temperature and pressure. The developed blanket contains a 24 cm of beryllium and 6 cm of solid breeder both with a 0.8 density factor. This blanket provides a local tritium breeding ratio of ∼2.0. The water coolant is isolated from the breeder material by several zones which eliminates the tritium buildup in the water by permeation and reduces the changes for water-breeder interaction. This improves the safety and environmental aspects of the blanket and eliminates the costly process of the tritium recovery from the water. 12 refs., 13 tabs

Tritium transport analysis for CFETR WCSB blanket

Energy Technology Data Exchange (ETDEWEB)

Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Yang, Wanli; Li, Yuanjie; Ge, Zhihao; Nie, Xingchen; Gao, Zhongping

2017-01-15

Highlights: • A simplified tritium transport model for CFETR WCSB blanket was developed. • Tritium transport process in CFETR WCSB blanket was analyzed. • Sensitivity analyses of tritium transport parameters were carried out. - Abstract: Water Cooled Solid Breeder (WCSB) blanket was put forward as one of the breeding blanket candidate schemes for Chinese Fusion Engineering Test Reactor (CFETR). In this study, a simplified tritium transport model was developed. Based on the conceptual engineering design, neutronics and thermal-hydraulic analyses of CFETR WCSB blanket, tritium transport process was analyzed. The results show that high tritium concentration and inventory exist in primary water loop and total tritium losses exceed CFETR limits under current conditions. Conducted were sensitivity analyses of influential parameters, including tritium source, temperature, flow-rate capacity and surface condition. Tritium performance of WCSB blanket can be significantly improved under a smaller tritium impinging rate, a larger flow-rate capacity or a better surface condition. This work provides valuable reference for the enhancement of tritium transport behavior in CFETR WCSB blanket.

Blanket comparison and selection study. Volume I

International Nuclear Information System (INIS)

1983-10-01

The objectives of the Blanket Comparison and Selection Study (BCSS) can be stated as follows: (1) Define a small number (approx. 3) of blanket design concepts that should be the focus of the blanket R and D program. A design concept is defined by the selection of all materials (e.g., breeder, coolant, structure and multiplier) and other major characteristics that significantly influence the R and D requirements. (2) Identify and prioritize the critical issues for the leading blanket concepts. (3) Provide the technical input necessary to develop a blanket R and D program plan. Guidelines for prioritizing the R and D requirements include: (a) critical feasibility issues for the leading blanket concepts will receive the highest priority, and (b) for equally important feasibility issues, higher R and D priority will be given to those that require minimum cost and short time

Fusion-Fission Transmutation Scheme-Efficient destruction of nuclear waste

International Nuclear Information System (INIS)

Kotschenreuther, M.; Valanju, P.M.; Mahajan, S.M.; Schneider, E.A.

2009-01-01

A fusion-assisted transmutation system for the destruction of transuranic nuclear waste is developed by combining a subcritical fusion-fission hybrid assembly uniquely equipped to burn the worst thermal nonfissile transuranic isotopes with a new fuel cycle that uses cheaper light water reactors for most of the transmutation. The center piece of this fuel cycle, the high power density compact fusion neutron source (100 MW, outer radius <3 m), is made possible by a new divertor with a heat-handling capacity five times that of the standard alternative. The number of hybrids needed to destroy a given amount of waste is an order of magnitude below the corresponding number of critical fast-spectrum reactors (FRs) as the latter cannot fully exploit the new fuel cycle. Also, the time needed for 99% transuranic waste destruction reduces from centuries (with FR) to decades

Optimal Neutron Source and Beam Shaping Assembly for Boron Neutron Capture Therapy

International Nuclear Information System (INIS)

Vujic, J.; Greenspan, E.; Kastenber, W.E.; Karni, Y.; Regev, D.; Verbeke, J.M.; Leung, K.N.; Chivers, D.; Guess, S.; Kim, L.; Waldron, W.; Zhu, Y.

2003-01-01

There were three objectives to this project: (1) The development of the 2-D Swan code for the optimization of the nuclear design of facilities for medical applications of radiation, radiation shields, blankets of accelerator-driven systems, fusion facilities, etc. (2) Identification of the maximum beam quality that can be obtained for Boron Neutron Capture Therapy (BNCT) from different reactor-, and accelerator-based neutron sources. The optimal beam-shaping assembly (BSA) design for each neutron source was also to e obtained. (3) Feasibility assessment of a new neutron source for NCT and other medical and industrial applications. This source consists of a state-of-the-art proton or deuteron accelerator driving and inherently safe, proliferation resistant, small subcritical fission assembly

Trade-off study of liquid metal self-cooled blankets

International Nuclear Information System (INIS)

Gohar, Y.

1986-01-01

A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of this study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. The primary results of the study are as follows: a) the lithium-lead blanket achieves a higher TBR with a smaller blanket thickness relative to the lithium blanket; b) the lithium blanket generates more energy per fusion neutron relative to the lithium-lead blanket; c) among the possible reflector materials, the carbon reflector produces the highest TBR; d) the high-Z reflector materials (Mo, Cu, W, or steel) generate more energy per fusion neutron and produce smaller TBRs relative to the carbon reflector; e) lithium-6 enrichment is required for the lithium-lead blanket to reduce the total blanket thickness; and f) the energy deposition per fusion neutron reaches a saturation as the blanket thickness, the fraction of the high-Z material in the reflector, or the reflector zone thickness increases (this allows one to design the blanket for a specific TBR without reducing the energy production)

Workshop on cold-blanket research

International Nuclear Information System (INIS)

1977-05-01

The objective of the workshop was to identify and discuss cold-plasma blanket systems. In order to minimize the bombardment of the walls by hot neutrals the plasma should be impermeable. This requires a density edge-thickness product of nΔ > 10 15 cm -2 . An impermeable cold plasma-gas blanket surrounding a hot plasma core reduces the plasma wall/limiter interaction. Accumulation of impurities in this blanket can be expected. Fuelling from a blanket may be possible as shown by experimental results, though not fully explained by classical transport of neutrals. Refuelling of a reacting plasma had to be ensured by inward diffusion. Experimental studies of a cold impermeable plasma have been done on the tokamak-like Ringboog device. Simulation calculations for the next generation of large tokamaks using a particular transport model, indicate that the plasma edge profile can be controlled to reduce the production of sputtered impurities to an acceptable level. Impurity control requires a small fraction of the radial space to accomodate the cold-plasma layer. The problem of exhaust is, however, more complicated. If the cold-blanket scheme works as predicted in the model calculations, then α-particles generated by fusion will be transported to the cold outside layer. The Communities' experimental programme of research has been discussed in terms of the tokamaks which are available and planned. Two options present themselves for the continuation of cold-blanket research

The blanket interface to TSTA

International Nuclear Information System (INIS)

Clemmer, R.G.; Finn, P.A.; Grimm, T.L.; Sze, D.K.; Anderson, J.L.; Bartlit, J.R.; Naruse, Y.; Yoshida, H.

1988-01-01

The requirements of tritium technology are centered in three main areas, (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The Tritium Systems Test Assembly (TSTA) now in operation at Los Alamos National Laboratory (LANL) is dedicated to developing and demonstrating the tritium technology for fuel processing and containment. TSTA is the only fusion fuel processing facility that can operate in a continuous closed-loop mode. The tritium throughput of TSTA is 1000 g/d. However, TSTA does not have a blanket interface system. The authors have initiated a study to define a Breeder Blanket Interface (BBIO) for TSTA. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. Various methods of tritium recovery from liquid lithium were assessed: yttrium gettering, permeation windows, and molten salt extraction. The authors' evaluation concluded that the best method was molten salt extraction

Characterizing subcritical assemblies with time of flight fixed by energy estimation distributions

Science.gov (United States)

Monterial, Mateusz; Marleau, Peter; Pozzi, Sara

2018-04-01

We present the Time of Flight Fixed by Energy Estimation (TOFFEE) as a measure of the fission chain dynamics in subcritical assemblies. TOFFEE is the time between correlated gamma rays and neutrons, subtracted by the estimated travel time of the incident neutron from its proton recoil. The measured subcritical assembly was the BeRP ball, a 4.482 kg sphere of α-phase weapons grade plutonium metal, which came in five configurations: bare, 0.5, 1, and 1.5 in iron, and 1 in nickel closed fitting shell reflectors. We extend the measurement with MCNPX-PoliMi simulations of shells ranging up to 6 inches in thickness, and two new reflector materials: aluminum and tungsten. We also simulated the BeRP ball with different masses ranging from 1 to 8 kg. A two-region and single-region point kinetics models were used to model the behavior of the positive side of the TOFFEE distribution from 0 to 100 ns. The single region model of the bare cases gave positive linear correlations between estimated and expected neutron decay constants and leakage multiplications. The two-region model provided a way to estimate neutron multiplication for the reflected cases, which correlated positively with expected multiplication, but the nature of the correlation (sub or superlinear) changed between material types. Finally, we found that the areal density of the reflector shells had a linear correlation with the integral of the two-region model fit. Therefore, we expect that with knowledge of reflector composition, one could determine the shell thickness, or vice versa. Furthermore, up to a certain amount and thickness of the reflector, the two-region model provides a way of distinguishing bare and reflected plutonium assemblies.

Activity report of working party on reactor physics of subcritical system. October 2001 to March 2003

International Nuclear Information System (INIS)

2004-03-01

Under the Research Committee on Reactor Physics, the Working Party on Reactor Physics of Subcritical System (ADS-WP) was set in July 2001 to research reactor physics of subcritical system such as Accelerator-Driven System (ADS). The WP, at the first meeting, discussed a guideline of its activity for two years and decided to perform theoretical research for the following subjects: (1) study of reactor physics for a subcritical core, (2) benchmark problems for a subcritical core and their calculations, (3) study of physical parameters affecting to set subcriticality of ADS, and (4) study of measurement and surveillance methods of subcriticality of a subcritical core. The activity of ADS-WP continued up to March 2003. In this duration, the members of the WP met together eight times, including four meetings jointly held with the Workshop on Accelerator-Driven Subcritical Reactor at Kyoto University Research Reactor Institute. This report summarizes the result obtained by the above WP activity and research. (author)

Convertible shielding to ceramic breeding blanket

International Nuclear Information System (INIS)

Furuya, Kazuyuki; Kurasawa, Toshimasa; Sato, Satoshi; Nakahira, Masataka; Togami, Ikuhide; Hashimoto, Toshiyuki; Takatsu, Hideyuki; Kuroda, Toshimasa.

1995-05-01

Four concepts have been studied for the ITER convertible blanket: 1)Layered concept 2)BIT(Breeder-Inside-Tube)concept 3)BOT(Breeder-Out of-Tube)concept 4)BOT/mixed concept. All concepts use ceramic breeder and beryllium neutron multiplier, both in the shape of small spherical pebbles, 316SS structure, and H 2 O coolant (inlet/outlet temperatures : 100/150degC, pressure : 2 MPa). During the BPP, only beryllium pebbles (the primary pebble in case of BOT/mixed concept) are filled in the blanket for shielding purpose. Then, before the EPP operation, breeder pebbles will be additionally inserted into the blanket. Among possible conversion methods, wet method by liquid flow seems expecting for high and homogeneous pebble packing. Preliminary 1-D neutronics calculation shows that the BOT/mixed concept has the highest breeding and shielding performance. However, final selection should be done by R and D's and more detail investigation on blanket characteristics and fabricability. Required R and D's are also listed. With these efforts, the convertible blanket can be developed. However, the following should be noted. Though many of above R and D's are also necessary even for non-convertible blanket, R and D's on convertibility will be one of the most difficult parts and need significant efforts. Besides the installation of convertible blanket with required structures and lines for conversion will make the ITER basic machine more complicated. (author)

Classification Using Markov Blanket for Feature Selection

DEFF Research Database (Denmark)

Zeng, Yifeng; Luo, Jian

2009-01-01

Selecting relevant features is in demand when a large data set is of interest in a classification task. It produces a tractable number of features that are sufficient and possibly improve the classification performance. This paper studies a statistical method of Markov blanket induction algorithm...... for filtering features and then applies a classifier using the Markov blanket predictors. The Markov blanket contains a minimal subset of relevant features that yields optimal classification performance. We experimentally demonstrate the improved performance of several classifiers using a Markov blanket...... induction as a feature selection method. In addition, we point out an important assumption behind the Markov blanket induction algorithm and show its effect on the classification performance....

Some neutronics of innovative subcritical assembly with fast neutron spectrum

International Nuclear Information System (INIS)

Kiyavitskaya, H.; Fokov, Yu.; Rutkovskaya, Ch.; Sadovich, S.; Kasuk, D.; Gohar, Y.; Bolshinsky, I.

2013-01-01

Conclusion: • New assembly can be used to: • develop the experimental techniques and adapt the existing ones for monitoring the sub-criticality level, neutron spectra measurements, etc; • study the spatial kinetics of sub-critical and critical systems with fast neutron spectra; • measure the transmutation reaction rates of minor-actinides etc

Test Blanket Working Group's recent activities

International Nuclear Information System (INIS)

Vetter, J.E.

2001-01-01

The ITER Test Blanket Working Group (TBWG) has continued its activities during the period of extension of the EDA with a revised charter on the co-ordination of the development work performed by the Parties and by the JCT leading to a co-ordinated test programme on ITER for a DEMO-relevant tritium breeding blanket. This follows earlier work carried out until July 1998, which formed part of the ITER Final Design Report (FDR), completed in 1998. Whilst the machine parameters for ITER-FEAT have been significantly revised compared to the FDR, testing of breeding blanket modules remains a main objective of the test programme and the development of a reactor-relevant breeding blanket to ensure tritium fuel self-sufficiency is recognized a key issue for fusion. Design work and R and D on breeding blanket concepts, including co-operation with the other Contacting Parties of the ITER-EDA for testing these concepts in ITER, are included in the work plans of the Parties

Achievements of element technology development for breeding blanket

International Nuclear Information System (INIS)

Enoeda, Mikio

2005-03-01

Japan Atomic Energy Research Institute (JAERI) has been performing the development of breeding blanket for fusion power plant, as a leading institute of the development of solid breeder blankets, according to the long-term R and D program of the blanket development established by the Fusion Council of Japan in 1999. This report is an overview of development plan, achievements of element technology development and future prospect and plan of the development of the solid breeding blanket in JAERI. In this report, the mission of the blanket development activity in JAERI, key issues and roadmap of the blanket development have been clarified. Then, achievements of the element technology development were summarized and showed that the development has progressed to enter the engineering testing phase. The specific development target and plan were clarified with bright prospect. Realization of the engineering test phase R and D and completion of ITER test blanket module testing program, with universities/NIFS cooperation, are most important steps in the development of breeding blanket of fusion power demonstration plant. (author)

Investigation of tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2

International Nuclear Information System (INIS)

Yildiz, K.; Sahin, S.; Sahin, H. M.; Acir, A.; Yalcin, S.; Altinok, T.; Bayrak, M.; Alkan, M.; Durukan, O.

2007-01-01

In the world, thorium reserves are three times more than natural Uranium reserves. It is planned in the near future that nuclear reactors will use thorium as a fuel. Thorium is not a fissile isotope because it doesn't make fission with thermal neutrons so it could be converted to 2 33U isotope which has very high quality fission cross-section with thermal neutrons. 2 33U isotope can be used in present LWRs as an enrichment fuel. In the fusion reactors, tritium is the most important fossil fuel. Because tritium is not natural isotope, it has to be produced in the reactor. The purpose of this work is to investigate the tritium and 2 33U breeding in a fission-fusion hybrid reactor fuelling with ThO 2 for Δt=10 days during a reactor operation period in five years. The neutronic analysis is performed on an experimental hybrid blanket geometry. In the center of the hybrid blanket, there is a line neutron source in a cylindrical cavity, which simulates the fusion plasma chamber where high energy neutrons (14.1 MeV) are produced. The conventional fusion reaction delivers the external neutron source for blankets following, 2 D + 3 T →? 4 He (3.5 MeV) + n (14.1 MeV). (1) The fuel zone made up of natural-ThO 2 fuel and it is cooled with different coolants. In this work, five different moderator materials, which are Li 2 BeF 4 , LiF-NaF-BeF 2 , Li 2 0Sn 8 0, natural Lithium and Li 1 7Pb 8 3, are used as coolants. The radial reflector, called tritium breeding zones, is made of different Lithium compounds and graphite in sandwich structure. In the work, eight different Lithium compounds were used as tritium breeders in the tritium breeding zones, which are Li 3 N, Li 2 O, Li 2 O 2 , Li 2 TiO 3 , Li 4 SiO 3 , Li 2 ZrO 3 , LiBr and LiH. Neutron transport calculations are conducted in spherical geometry with the help of SCALE4.4A SYSTEM by solving the Boltzmann transport equation with code CSAS and XSDRNPM, under consideration of unresolved and resolved resonances, in S 8 -P 3

Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

International Nuclear Information System (INIS)

Panayotov, Dobromir; Grief, Andrew; Merrill, Brad J.; Humrickhouse, Paul; Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon; Poitevin, Yves; Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard

2016-01-01

Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

Energy Technology Data Exchange (ETDEWEB)

Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Grief, Andrew [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Merrill, Brad J.; Humrickhouse, Paul [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID (United States); Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom)

2016-11-01

Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

Extraction of antioxidants from Chlorella sp. using subcritical water treatment

Science.gov (United States)

Zakaria, S. M.; Mustapa Kamal, S. M.; Harun, M. R.; Omar, R.; Siajam, S. I.

2017-06-01

Chlorella sp. microalgae is one of the main source of natural bioactive compounds used in the food and pharmaceutical industries. Subcritical water extraction is the technique that offers an efficient, non-toxic, and environmental-friendly method to obtain natural ingredients. In this work, the extracts of Chlorella sp. microalgae was evaluated in terms of: chemical composition, extraction (polysaccharides) yield and antioxidant activity, using subcritical water extraction. Extractions were performed at temperatures ranging from 100°C to 300°C. The results show that by using subcritical water, the highest yield of polysaccharides is 23.6 that obtained at 150°C. Analysis on the polysaccharides yield show that the contents were highly influenced by the extraction temperature. The individual antioxidant activity were evaluated by in vitro assay using a free radical method. In general, the antioxidant activity of the extracts obtained at different water temperatures was high, with values of 31.08-54.29 . The results indicated that extraction by subcritical water was effective and Chlorella sp. can be a useful source of natural antioxidants.

Design study of blanket structure for tokamak experimental fusion reactor

International Nuclear Information System (INIS)

1979-11-01

Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

Physics study of D-D/D-T neutron driven experimental subcritical assembly

International Nuclear Information System (INIS)

Sinha, Amar

2015-01-01

An experimental program to design and study external source driven subcritical assembly has been initiated at BARC. This program is aimed at understanding neutronic characteristics of accelerator driven system at low power level. In this series, a zero-power, sub-critical assembly driven by a D-D/D-T neutron generator has been developed. This system is modular in design and it is first in the series of subcritical assemblies being designed. The subcritical core consists of natural uranium fuel with high density polyethylene as moderator and beryllium oxide as reflector. The subcritical core is coupled to Purnima Neutron Generator. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques. Further experiments are being planned to measure the reactivity and other kinetic parameters using noise methods. This facility would also be used for carrying out studies on effect of source importance and measurement of source multiplication factor k s and external neutron source efficiency φ* in great details. Some experiments with D-D and D-T neutrons have been presented. (author)

ITER breeding blanket module design and analysis

Energy Technology Data Exchange (ETDEWEB)

Kuroda, Toshimasa; Enoeda, Mikio; Kikuchi, Shigeto [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1998-11-01

The ITER breeding blanket employs a ceramic breeder and Be neutron multiplier both in small spherical pebble form. Radial-poloidal cooling panels are arranged in the blanket box to remove the nuclear heating in these materials and to reinforce the blanket structure. At the first wall, Be armor is bonded onto the stainless steel (SS) structure to provide a low Z plasma-compatible surface and to protect the first wall/blanket structure from the direct contact with the plasma during off-normal events. Thermo-mechanical analyses and investigation of fabrication procedure have been performed for this breeding blanket. To evaluate thermo-mechanical behavior of the pebble beds including the dependency of the effective thermal conductivity on stress, analysis methods have been preliminary established by the use of special calculation option of ABAQUS code, which are briefly summarized in this report. The structural response of the breeding blanket module under internal pressure of 4 MPa (in case of in-blanket LOCA) resulted in rather high stress in the blanket side (toroidal end) wall, thus addition of a stiffening rib or increase of the wall thickness will be needed. Two-dimensional elasto-plastic analyses have been performed for the Be/SS bonded interface at the first wall taking a fabrication process based on HIP bonding and thermal cycle due to pulsed plasma operation into account. The stress-strain hysteresis during these process and operation was clarified, and a procedure to assess and/or confirm the bonding integrity was also proposed. Fabrication sequence of the breeding blanket module was preliminarily developed based on the procedure to fabricate part by part and to assemble them one by one. (author)

Modeling of Parameters of Subcritical Assembly SAD

CERN Document Server

Petrochenkov, S; Puzynin, I

2005-01-01

The accepted conceptual design of the experimental Subcritical Assembly in Dubna (SAD) is based on the MOX core with a nominal unit capacity of 25 kW (thermal). This corresponds to the multiplication coefficient $k_{\\rm eff} =0.95$ and accelerator beam power 1 kW. A subcritical assembly driven with the existing 660 MeV proton accelerator at the Joint Institute for Nuclear Research has been modelled in order to make choice of the optimal parameters for the future experiments. The Monte Carlo method was used to simulate neutron spectra, energy deposition and doses calculations. Some of the calculation results are presented in the paper.

Survey on the fusion/fission-hybrid-reactors, a literature review

International Nuclear Information System (INIS)

A survey, based on existing literature, of the work being pursued worldwide on fusion - fission (hybrid) reactor systems is presented. Six areas are reviewed: Plasma physics parameters; Blankets concepts; Fuel cycles; Reactor conceptual designs; Safety and environmental problems; System studies and economic perspectives. Attention has been restricted to systems using magnetically confined plasmas, mainly to mirror and Tokamak - type concepts. The aim is to provide sufficient information, even if not exhaustive, on hybrid reactor concepts in order to help understand what may be expected from their possible development and the ways in which hybrids could affect the future energy scenario. Some concluding remarks are made which represent the personal view of the authors only

Evaluating Subcriticality during the Ebola Epidemic in West Africa.

Directory of Open Access Journals (Sweden)

Wayne T A Enanoria

Full Text Available The 2014-2015 Ebola outbreak is the largest and most widespread to date. In order to estimate ongoing transmission in the affected countries, we estimated the weekly average number of secondary cases caused by one individual infected with Ebola throughout the infectious period for each affected West African country using a stochastic hidden Markov model fitted to case data from the World Health Organization. If the average number of infections caused by one Ebola infection is less than 1.0, the epidemic is subcritical and cannot sustain itself. The epidemics in Liberia and Sierra Leone have approached subcriticality at some point during the epidemic; the epidemic in Guinea is ongoing with no evidence that it is subcritical. Response efforts to control the epidemic should continue in order to eliminate Ebola cases in West Africa.

Methods for the reactivity evaluation in subcritical systems analysis: a review

International Nuclear Information System (INIS)

Dulla, S.; Picca, P.; Carta, M.

2011-01-01

The assessment of the subcritical source-driven system technology for waste incineration and power production requires the development of reliable and efficient techniques for the reactivity evaluation and monitoring. Starting from the standard methods developed for close-to-criticality systems, extensive research activities have been carried out to analyze the behavior of subcritical assembly in time-dependent condition and to infer the subcriticality level from local flux values. In the present work, a review of some key aspects in the method development for ADS analysis is proposed, with special attention to the techniques for reactivity evaluation. (author)

APT target-blanket fabrication development

Energy Technology Data Exchange (ETDEWEB)

Fisher, D.L.

1997-06-13

Concepts for producing tritium in an accelerator were translated into hardware for engineering studies of tritium generation, heat transfer, and effects of proton-neutron flux on materials. Small-scale target- blanket assemblies were fabricated and material samples prepared for these performance tests. Blanket assemblies utilize composite aluminum-lead modules, the two primary materials of the blanket. Several approaches are being investigated to produce large-scale assemblies, developing fabrication and assembly methods for their commercial manufacture. Small-scale target-blanket assemblies, designed and fabricated at the Savannah River Site, were place in Los Alamos Neutron Science Center (LANSCE) for irradiation. They were subjected to neutron flux for nine months during 1996-97. Coincident with this test was the development of production methods for large- scale modules. Increasing module size presented challenges that required new methods to be developed for fabrication and assembly. After development, these methods were demonstrated by fabricating and assembling two production-scale modules.

Analysis of neutronics and dynamic characteristics with reactivity injection in LBE cooled sub-critical reactor

International Nuclear Information System (INIS)

Chen Sen; Wu Yican; Jin Ming; Chen Zhibin; Bai Yunqing; Zhao Zhumin

2014-01-01

Accelerator Driven Sub-critical System (ADS) has particular neutronics behaviors compared with the critical system. Prompt jump approximation point reactor kinetics equations taken external source into account have been deduced using an approach of prompt jump approximation. And the relationship between injection reactivity and power ampliation has been achieved. In addition, based on the RELAP5 code the prolong development of point reactor kinetics code used into assessing sub-critical system have been promoted. Different sub-criticality (k eff = 0.90, 0.95, 0.97, 0.98 and 0.99) have been assessed in preliminary design of a type of natural circulation cooling sub-critical reactor under conditions of reactivity injection +1 β in one second. It shows that the external source prompt transient approximation method has an accurate solution after injecting reactivity around short time and has a capacity to solve the dynamic equation, and the sub-critical system has an inner stability while the deeper sub-criticality the less impact on the sub-critical system. (authors)

Tokamak hybrid thermonuclear reactor for the production of fissionable fuel and electric power

International Nuclear Information System (INIS)

Velikhov, E.P.; Glukhikh, V.A.; Gur'ev, V.V.

1978-01-01

The results of feasibility studies of a tokamak- based hybrid reactor concept are presented. The system selected has a D-T plasma volume of 575 m 3 with additional plasma heating by injection of fast neutral particles. The method of heating makes it possible to achieve an economical two-component tokamak regime at ntau=(4-6)x10 13 sxcm -3 , i e. far below the Lawson criterion. Plasma and vacuum chamber are surrounded by a blanket where fissionable plutonium is produced and heat transformed into electric power is generated. Major plasma-neutron-physical characteristics of the 6905 MWth (2500 MWe) reactor and its electromagnetic system are presented. Evaluations show that the hybrid reactor can produce about 800 kg of Pu per 1GWth/yr as compared to 70-150 kg of Pu for fast breeder reactors. The increased Pu production rate is the major merit of the concept promising for both power generation and fuelling thermal fission reactions

On blanket concepts of the Helias reactor

International Nuclear Information System (INIS)

Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

1999-07-01

The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m 2 ) the average neutron power load on the first wall is below 1 MWm .2 , which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

Choosing the optimal parameters of subcritical reactors driven by accelerators

International Nuclear Information System (INIS)

Khudaverdyan, A.G.; Zhamkochyan, V.M.

1998-03-01

Physical aspects of a subcritical Nuclear Power Plants (NPP) driven by proton accelerators are considered. Estimated theoretical calculations are made for subcritical regimes of various types of reactors. It was shown that the creation of the quite effective explosion-safe NPP is real at an existing level of the accelerator technique by using available reactor units (including the serial ones). (author)

Magnetoconvection in HCLL blankets

International Nuclear Information System (INIS)

Mistrangelo, C.; Buehler, L.

2014-01-01

In the present work we consider magneto-convective flows in one of the proposed European liquid metal blankets that will be tested in the experimental fusion reactor ITER. Here the PbLi alloy is used as breeder material and helium as coolant. In order to finalize the design of the helium cooled lead lithium (HCLL) blanket, studies are still required to fully understand the behavior of the electrically conducting breeder under the influence of the intense magnetic field that confines the fusion plasma and in case of non-uniform thermal conditions. Liquid metal HCLL blanket flows are expected to be mainly driven by buoyancy forces caused by non-isothermal operating conditions due to neutron volumetric heating and cooling of walls, since only a weak forced ow is foreseen for tritium extraction in external ancillary systems. Buoyancy can therefore become very important and modify the velocity distribution and related heat transfer performance of the blanket. The present numerical study aims at clarifying the influence of electromagnetic and thermal coupling of neighboring fluid domains on magneto-convective flows in geometries relevant for the HCLL blanket concept. According to the last design review two internal cooling plates subdivide the fluid domain into three slender flow regions, which are thermally and electrically coupled through common walls. First a uniform volumetric heat source is considered to identify the basic convective patterns that establish in the liquid metal. Results are then compared with those obtained by applying a realistic radial distribution of the power density as obtained from a neutronic analysis. Velocity and temperature distributions are discussed for various volumetric heat sources and magnetic field strengths.

Benchmark calculations for fusion blanket development

International Nuclear Information System (INIS)

Sawan, M.E.; Cheng, E.T.

1985-01-01

Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li/sub 17/Pb/sub 83/ and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the TBR to group structure and weighting spectrum increases and Li enrichment decrease with up to 20% discrepancies for thin natural Li/sub 17/Pb/sub 83/ blankets

Transient subcritical crack-growth behavior in transformation-toughened ceramics

International Nuclear Information System (INIS)

Dauskardt, R.H.; Ritchie, R.O.; Carter, W.C.; Veirs, D.K.

1990-01-01

Transient subcritical crack-growth behavior following abrupt changes in the applied load are studied in transformation-toughened ceramics. A mechanics analysis is developed to model the transient nature of transformation shielding of the crack tip, K s , with subcritical crack extension following the applied load change. conditions for continued crack growth, crack growth followed by arrest, and no crack growth after the load change, are considered and related to the magnitude and sign of the applied load change and to materials properties such as the critical transformation stress. The analysis is found to provide similar trends in K s compared to values calculated from experimentally measured transformation zones in a transformation-toughened Mg-PSZ. In addition, accurate prediction of the post load-change transient crack-growth behavior is obtained using experimentally derived steady-state subcritical crack-growth relationships for cyclic fatigue in the same material

Stress analysis of the tokamak engineering test breeder blanket

International Nuclear Information System (INIS)

Huang Zhongqi

1992-01-01

The design features of the hybrid reactor blanket and main parameters are presented. The stress analysis is performed by using computer codes SAP5p and SAP6 with the three kinds of blanket module loadings, i.e, the pressure of coolant, the blanket weight and the thermal loading. Numerical calculation results indicate that the stresses of the blanket are smaller than the allowable ones of the material, the blanket design is therefore reasonable

NET test blanket design and remote maintenance

International Nuclear Information System (INIS)

Holloway, C.; Hubert, P.

1991-01-01

The NET machine has three horizontal ports reserved for testing tritium breeding blanket designs during the physics phase and possibly five during the technology phase. The design of the ports and test blankets are modular to accept a range of blanket options, provide radiation shielding and allow routine replacement. Radiation levels during replacement or maintenance require that all operations must be carried out remotely. The paper describes the problems overcome in providing a port design which includes attachment to the vacuum vessel with double vacuum seals, an integrated cooled first wall and support guides for the test blanket module. The method selected to remotely replace the test module whilst controlling the spread of contamination is also adressed. The paper concludes that the provisions of a test blanket facility based on the NET machine design is feasible. (orig.)

Remote handling demonstration of ITER blanket module replacement

International Nuclear Information System (INIS)

Kakudate, S.; Nakahira, M.; Oka, K.; Taguchi, K.; Obara, K.; Tada, E.; Shibanuma, K.; Tesini, A.; Haange, R.; Maisonnier, D.

2001-01-01

In ITER, the in-vessel components such as blanket are to be maintained or replaced remotely since they will be activated by 14 MeV neutrons, and a complete exchange of shielding blanket with breeding blanket is foreseen after the Basic Performance Phase. The blanket is segmented into about seven hundred modules to facilitate remote maintainability and allow individual module replacement. For this, the remote handing equipment for blanket maintenance is required to handle a module with a dead weight of about 4 tonne within a positioning accuracy of a few mm under intense gamma radiation. According to the ITER R and D program, a rail-mounted vehicle manipulator system was developed and the basic feasibility of this system was verified through prototype testing. Following this, development of full-scale remote handling equipment has been conducted as one of the ITER Seven R and D Projects aiming at a remote handling demonstration of the ITER blanket. As a result, the Blanket Test Platform (BTP) composed of the full-scale remote handling equipment has been completed and the first integrated performance test in March 1998 has shown that the fabricate remote handling equipment satisfies the main requirements of ITER blanket maintenance. (author)

An assessment of the base blanket for ITER

International Nuclear Information System (INIS)

Raffray, A.R.; Abdou, M.A.; Ying, A.

1991-01-01

Ideally, the ITER base blanket would provide the necessary tritium for the reactor to be self-sufficient during operation, while having minimal impact on the overall reactor cost, reliability and safety. A solid breeder blanket has been developed in CDA phase in an attempt to achieve such objectives. The reference solid breeder base blanket configurations at the end of the CDA phase has many attractive features such as a tritium breeding ratio (TBR) of 0.8--0.9 and a reasonably low tritium inventory. However, some concerns regarding the risk, cost and benefit of the base blanket have been raised. These include uncertainties associated with the solid breeder thermal control and the potentially high cost of the amount of Be used to achieve high TBR and to provide the necessary thermal barrier between the high temperature solid breeder and low temperature coolant. This work addresses these concerns. The basis for the selection of a breeding blanket is first discussed in light of the incremental risk, cost and benefits relative to a non-breeding blanket. Key issues associated with the CDA breeding blanket configurations are then analyzed. Finally, alternative schemes that could enhance the attractiveness and flexibility of a breeding blanket are explored

Design analyses of self-cooled liquid metal blankets

International Nuclear Information System (INIS)

Gohar, Y.

1986-12-01

A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, a study was carried out to assess the impact of different reactor design choices on the reactor performance parameters. The design choices include the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, and the coolant choice for the nonbreeding inboard blanket. In addition, tritium breeding benchmark calculations were performed using different transport codes and nuclear data libraries. The importance of the TBR in the blanket design motivated the benchmark calculations

Preliminary study on lithium-salt aqueous solution blanket

International Nuclear Information System (INIS)

Yoshida, Hiroshi; Naruse, Yuji; Yamaoka, Mitsuaki; Ohara, Atsushi; Ono, Kiyoshi; Kobayashi, Shigetada.

1992-06-01

Aqueous solution blanket using lithium salts such as LiNO 3 and LiOH have been studied in the US-TIBER program and ITER conceptual design activity. In the JAERI/LANL collaboration program for the joint operation of TSTA (Tritium Systems Test Assembly), preliminary design work of blanket tritium system for lithium ceramic blanket, aqueous solution blanket and liquid metal blanket, have been performed to investigate technical feasibility of tritium demonstration tests using the TSTA. Detail study of the aqueous solution blanket concept have not been performed in the Japanese fusion program, so that this study was carried out to investigate features of its concept and to evaluated its technical problems. The following are the major items studied in the present work: (i) Neutronics of tritium breeding ratio and shielding performance Lithium concentration, Li-60 enrichment, beryllium or lead, composition of structural material/beryllium/solution, heavy water, different lithium-salts (ii) Physicochemical properties of salts Solubility, corrosion characteristics and compatibility with structural materials, radiolysis (iii) Estimation of radiolysis in ITER aqueous solution blanket. (author)

Nuclear characteristics of D-D fusion reactor blankets

International Nuclear Information System (INIS)

Nakashima, Hideki; Ohta, Masao

1978-01-01

Fusion reactors operating on deuterium (D-D) cycle are considered to be of long range interest for their freedom from tritium breeding in the blanket. The present paper discusses the various possibilities of D-D fusion reactor blanket designs mainly from the standpoint of the nuclear characteristics. Neutronic and photonic calculations are based on presently available data to provide a basis of the optimal blanket design in D-D fusion reactors. It is found that it appears desirable to design a blanket with blanket/shield (BS) concept in D-D fusion reactors. The BS concept is designed to obtain reasonable shielding characteristics for superconducting magnet (SCM) by using shielding materials in the compact blanket. This concept will open the possibility of compact radiation shield design based on assured technology, and offer the advantage from the system economics point of view. (auth.)

ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

International Nuclear Information System (INIS)

WONG, CPC; MALANG, S; NISHIO, S; RAFFRAY, R; SAGARA, S

2002-01-01

OAK A271 ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS. First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability

Accelerator driven subcritical reactors

International Nuclear Information System (INIS)

Salvatores, M.

2001-01-01

ADS concepts have been proposed in the last decade for a variety of applications. However, there is a convergence of interest of several countries and laboratories on the application of ADS to transmutation. This applies to plutonium, and/or minor actinides (MA) and long-lived fission products (LLFP). As far as the so-called partitioning and transmutation (PIT) strategies, it was indicated that they can be clarified according to the option taken with respect to Pu and MA, i.e., a) keep Pu and MA together, b) separate Pu from MA. At present several programs are going on ADS: in Japan, USA Europe, where activities in 9 countries are coordinated by a European Technical Working Group (ETWG), and in Russia. As far as the implications for the definition of nuclear data needs, dedicated subcritical cores should have new type of fuels (Pu+MA in different proportions). Proposals are being worked out. For example, composite (such as ceramic-metallic or ceramic-ceramic) fuels are presently under study. The actinide oxide is dispersed in a metallic matrix (Zr, or W or Mo) or in an oxide matrix (e.g., MgO). In these cases, reliable data are required for the matrix materials. As far as coolants, Pb/Bi, Pb, and gas are considered, besides Na. Hard (or very hard) fast neutron spectrum is required. As far as LLFP, transmutation strategies in ADS are proposed. Candidates are 129 I, 99 Tc, 135 Cs, but also 79 Se, 107 Pd, 93 Zr etc. At present, there is no clear option for their transmutation (one needs a high level of thermalized neutrons, support matrixes for target irradiation, isotopic separations, reprocessing techniques, etc.). Finally, ADS transmutation will give rise to fuel cycles, where very active materials will be present. Cm and higher mass isotopes (up to 252 Cf) will be contributors to dose and neutron source strength. This area will deserve attention in future, in order to define the relevant data needs. It is recommended to coordinate work on MA data as a priority

Benchmark calculations for fusion blanket development

International Nuclear Information System (INIS)

Sawan, M.L.; Cheng, E.T.

1986-01-01

Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li 17 Pb 83 and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the tritium breeding ratio to group structure and weighting spectrum increases as the thickness and Li enrichment decrease with up to 20% discrepancies for thin natural Li 17 Pb 83 blankets. (author)

Linear accelerator fuel enricher regenerator (LAFER) and fission product transmutor (APEX)

International Nuclear Information System (INIS)

Steinberg, M.; Powell, J.R.; Takahashi, H.; Grand, P.; Kouts, H.J.C.

1979-01-01

In addition to safety, two other major problems face the nuclear industry today; first is the long-term supply of fissle material and second is the disposal of long-lived fission product waste. The higher energy proton linear accelerator can assist in the solution of each of these problems. High energy protons from the linear accelerator interact with a molten lead target to produce spallation and evaporation neutrons. The neutrons are absorbed in a surrounding blanket of light water power reactor (LWR) fuel elements to produce fissile Pu-239 or U-233 fuel from natural fertile U-238 or Th-232 contained in the elements. The fissile enriched fuel element is used in the LWR power reactor until its reactivity is reduced after which the element is regenerated in the linear accelerator target/blanket assembly and then the element is once again burned (fissioned) in the power LWR. In this manner the natural uranium fuel resource can supply an expanding nuclear power reactor economy without the need for fuel reprocessing, thus satisfying the US policy of non-proliferation. In addition, the quantity of spent fuel elements for long-term disposal is reduced in proportion to the number of fuel regeneration cycles through the accelerator. The limiting factor for in-situ regeneration is the burnup damage to the fuel cladding material. A 300 ma-1.5 GeV (450 MW) proton linear accelerator can produce approximately one ton of fissile (Pu-239) material annually which is enough to supply fuel to three 1000 MW(e) LWR power reactors. With two cycles of enriching and regenerating, the nuclear fuel natural resource can be stretched by a factor of 3.6 compared to present fuel cycle practice without the need for reprocessing. Furthermore, the need for isotopic enrichment facilities is drastically reduced

Breeding blankets for thermonuclear reactors

International Nuclear Information System (INIS)

Rocaboy, Alain.

1982-06-01

Materials with structures suitable for this purpose are studied. A bibliographic review of the main solid and liquid lithiated compounds is then presented. Erosion, dimensioning and maintenance problems associated with the limiter and the first wall of the reactor are studied from the point of view of the constraints they impose on the design of the blankets. Detailed studies of the main solid and liquid blanket concepts enable the best technological compromises to be determined for the indispensable functions of the blanket to be assured under acceptable conditions. Our analysis leads to four classes of solution, which cannot at this stage be considered as final recommendations, but which indicate what sort of solutions it is worthwhile exploring and comparing in order to be in a position to suggest a realistic blanket at the time when plasma control is sufficiently good for power reactors to be envisaged. Some considerations on the general architecture of the reactor are indicated. Energy storage with pulsed reactors is discussed in the appendix, and a first approach made to minimizing the total tritium recovery [fr

A comparison of fusion breeder/fission client and fission breeder/fission client systems for electrical energy production

International Nuclear Information System (INIS)

Land, R.J.; Parish, T.A.

1983-01-01

A parametric study that evaluated the economic performance of breeder/client systems is described. The linkage of the breeders to the clients was modelled using the stockpile approach to determine the system doubling time. Since the actual capital costs of the breeders are uncertain, a precise prediction of the cost of a breeder was not attempted. Instead, the breakeven capital cost of a breeder relative to the capital cost of a client reactor was established by equating the cost of electricity from the breeder/client system to the cost of a system consisting of clients alone. Specific results are presented for two breeder/client systems. The first consisted of an LMFBR with LWR clients. The second consisted of a DT fusion reactor (with a 238 U fission suppressed blanket) with LWR clients. The economics of each system was studied as a function of the cost of fissile fuel from a conventional source. Generally, the LMFBR/LWR system achieved relatively small breakeven capital cost ratios; the maximum ratio computed was 2.2 (achieved at approximately triple current conventional fissile material cost). The DTFR/LWR system attained a maximum breakeven capital cost ratio of 4.5 (achieved at the highest plasma quality (ignited device) and triple conventional fissile cost)

Tritium inventory and permeation in liquid breeder blankets

International Nuclear Information System (INIS)

Reiter, F.

1990-01-01

This report reviews studies of the transport of hydrogen isotopes in the DEMO relevant water-cooled Pb-17Li blanket to be tested in NET and in a self-cooled blanket which uses Pb-17Li or Flibe as a liquid breeder material and V or Fe as a first wall material. The time dependences of tritium inventory and permeation in these blankets and of deuterium and tritium recycling in the self-cooled blanket are presented and discussed

Characterization of the effects of continuous salt processing on the performance of molten salt fusion breeder blankets

International Nuclear Information System (INIS)

Patterson-Hine, F.A.; Davidson, J.W.; Klein, D.E.; Lee, J.D.

1985-01-01

Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs: fluorination only, fluorination plus reductive extraction, and fluorination, plus reductive extraction, plus metal transfer. The effects of processing on blanket performance have been assessed for these three levels of processing and various equilibrium uranium concentrations in the salt. A one-dimensional model of the blanket was used in the neutronics analysis, which incorporated transport calculations with time-dependent isotope generation and depletion calculations. The method of salt processing was found to have little affect on the level of radioactivity, toxicity, or the thermal behavior of the salt during operation of the reactor. The processing rates necessary to maintain the desired uranium concentrations in the suppressed-fission environment were quite low, which permitted only long-lived species to be removed from the salt. The effects of the processing therefore became apparent only after the radioactivity due to the short-lived species diminished. The effect of the additional processing (reductive extraction and metal transfer) could be seen after approximately 1 year of decay, but were not significant at times closer to shutdown. The reduced radioactivity and corresponding heat deposition were thus of no consequence in accident or maintenance situations. Net fissile production in the Be/MS blanket concept at a fusion power level of 3000 MW at 70% capacity ranged from 5100 kg/year to 5170 kg/year for uranium concentrations of 0.11% and 1.0% 233 U in thorium, respectively, with fluorination-only processing. The addition of processing by reductive extraction resulted in 5125 kg/year for the 0.11% 233 U case and 5225 kg/year for the 1.0% 233 U case

Solubility of Benzo[a]pyrene and Organic Matter of Soil in Subcritical Water

Directory of Open Access Journals (Sweden)

Svetlana Sushkova

2015-12-01

Full Text Available A dynamic subcritical water extraction method of benzo[a]pyrene from soils is under consideration. The optimum conditions for benzo[a]pyrene extraction from soil are described including the soil treatment by subcritical water at 250 °C and 100 atm for 30 min. The effectiveness of developed method was determined using the matrix spiking recovery technique. A comparative analysis was made to evaluate the results of benzo[a]pyrene extraction from soils using the subcritical water and organic solvents. The advantages of the subcritical water extraction involve the use of ecologically friendly solvent, a shorter time for the analysis and a higher amount of benzo[a]pyrene extracted from soil (96 %. The influence of subcritical water extraction on soil properties was measured the investigation of the processes occurring within soil under the influence the high temperature and pressure. Under appropriate conditions of the experiment there is the destruction of the soil organic matter while the composition of the soil mineral fraction remains practically unchanged.

Proposed sub-criticality level for an 80 MWTHd-bismuth-cooled Ads

International Nuclear Information System (INIS)

Mansani, L.; Monti, R.; Neuhold, P.

2003-01-01

The degree of operational sub-criticality of an Accelerator-driven System (ADS) on the one hand directly affects key accelerator system parameters, such as the proton beam current required to sustain the selected rated power level and, on the other, the likelihood of approaching or attaining criticality under abnormal or accident conditions. Then, if in all such conditions the safety goal is pursued to design the sub-critical core so that it stays away from criticality with adequate margin, the required operational sub-criticality level must be determined by a properly balanced approach between excessively demanding accelerator system performances and risk of accidental criticality. The approach must necessarily include evaluation and appropriate combination of the relevant reactivity effects (e.g. from system cool-down, postulated accident scenarios, geometrical variations) and proper consideration of specific design features (such as, for instance, the absence of safety rods, intended as neutron absorbing devices having a role equivalent to the shutdown rods in critical reactors). The paper presents a possible approach to the determination of the operational sub-criticality level of an 80 MWth Lead-Bismuth-cooled pool type ADS, initially conceived and developed by a team of Italian Organisations led by Ansaldo, with funding from the Ministry of University and Scientific and Technological Research, and currently in the process of being assessed, versus a gas-cooled concept, in the frame of a contract with the Commission of the European Communities. After a brief description of the Lead-Bismuth-cooled ADS concept relevant features and of the key safety goals in terms of required sub-criticality margin, the evaluated reactivity effects are presented, a method to combine them is discussed and a proposed operational sub-criticality level is derived. (author)

Objectives and status of EUROfusion DEMO blanket studies

Energy Technology Data Exchange (ETDEWEB)

Boccaccini, L.V., E-mail: lorenzo.boccaccini@kit.edu [Karlsruhe Institute of Technology (KIT) (Germany); Aiello, G.; Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Bachmann, C. [EUROfusion, PPPT, Garching (Germany); Barrett, T. [CCFE, Abingdon OX14 3DB (United Kingdom); Del Nevo, A. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Demange, D. [Karlsruhe Institute of Technology (KIT) (Germany); Forest, L. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Hernandez, F.; Norajitra, P. [Karlsruhe Institute of Technology (KIT) (Germany); Porempovic, G. [Fuziotech Engineering Ltd (Hungary); Rapisarda, D. [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Sardain, P. [CEA/IRFM, 13115 Saint-Paul-lès-Durance (France); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Vala, L. [Centrum výzkumu Řež, 250 68 Husinec-Řež (Czech Republic)

2016-11-01

Highlights: • Short description of the new Breeding Blanket Project in the EUROfusion consortium for the design of the EU PPPT DEMO: objectives. • Presentation of the design approach used in the development of the Breeding Blanket design: requirements. • Breeding Blanket design; in particular the four blanket concepts included in the study are presented, recent results highlighted and the status discussed. • Auxiliary systems and related R&D programme: in particular the work areas addressed in the Project (Tritium Technology, Pb-Li and Solid Breeders Technology, First Wall Design and R&D, Manufacturing) are presented, recent results highlighted and the status discussed. - Abstract: The design of a DEMO reactor requires the design of a blanket system suitable of reliable T production and heat extraction for electricity production. In the frame of the EUROfusion Consortium activities, the Breeding Blanket Project has been constituted in 2014 with the goal to develop concepts of Breeding Blankets for the EU PPPT DEMO; this includes an integrated design and R&D programme with the goal to select after 2020 concepts on fusion plants for the engineering phase. The design activities are presently focalized around a pool of solid and liquid breeder blanket with helium, water and PbLi cooling. Development of tritium extraction and control technology, as well manufacturing and development of solid and PbLi breeders are part of the programme.

Design project of fast subcritical system 'Mala Lasta'; Idejno resenje brzog podkriticnog sistema Mala LASTA

Energy Technology Data Exchange (ETDEWEB)

Milosevic, M; Stefanovic, D; Popovic, D; Pesic, M; Zavaljevski, N; Nikolic, D; Arsenovic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1988-10-15

This report contains two parts. Part one covers the objective and fundamental elements for the choice of fast subcritical system 'Mala Lasta', review of the existing fast subcritical assemblies, and a description of the available domestic computer codes applied for calculating neutron reactor parameters. Comparison of results obtained by these codes for a number of existing subcritical assemblies was used for the choice of the design project described in part two of this report. It contains detailed description of the operating parameters of the chosen subcritical system based on the obtained calculated parameters.

Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

International Nuclear Information System (INIS)

Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

1999-01-01

The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses

The trade experiment: Shielding calculations for the building hosting the subcritical system

International Nuclear Information System (INIS)

Burn, K. W.; Carta, M.; Casalini, L.; Kadi, Y.; Monti, S.; Nava, E.; Palomba, M.; Petrovich, C.; Picardi, L.; Rubbia, C.; Troiani, F.

2005-01-01

The TRADE project (TRIGA Accelerator Driven Experiment), to be performed at the existing TRIGA reactor at ENEA Casaccia, has been proposed as a validation of the accelerator-driven system (ADS) concept. TRADE will be the first experiment in which the three main components of an ADS-the accelerator, spallation target and sub-critical blanket-are coupled at a power level sufficient to encounter reactivity feedback effects. As such, TRADE represents the necessary intermediate step in the development of hybrid transmutation systems, its expected outcomes being considered crucial-in terms of proof of stability of operation, dynamic behaviour and licensing issues-for the subsequent realisation of an ADS Transmutation Demonstrator. An essential role in the feasibility study of the experiment is played by radioprotection calculations. Such a system exhibits new characteristics with respect to a traditional reactor, owing to the presence of the proton accelerator. As beam losses always occur under normal operating conditions of an accelerator, shielding studies need to be performed not only around the reactor but also along the beam line from the accelerator to the spallation target. This paper illustrates a preliminary evaluation, using Monte Carlo methods, of the additional shielding to be located around the reactor structures, the beam transport line and the existing reactor building to allow access into the reactor hall and to restrict the doses outside to their legal limits. (authors)

Design and development of ceramic breeder demo blanket

International Nuclear Information System (INIS)

Enoeda, M.; Sato, S.; Hatano, T.

2001-01-01

Ceramic breeder blanket development has been widely conducted in Japan from fundamental researches to project-oriented engineering scaled development. A long term R and D program has been launched in JAERI since 1996 as a course of DEMO blanket development. The objectives of this program are to provide engineering data base and fabrication technologies of the DEMO blanket, aiming at module testing in ITER currently scheduled to start from the beginning of the ITER operation as a near-term target. Two types of DEMO blanket systems, water cooled blanket and helium cooled blanket, have been designed to be consistent with the SSTR (Steady State Tokamak Reactor) which is the reference DEMO reactor design in JAERI. Both of them utilize packed small pebbles of breeder Li 2 O or Li 2 TiO 3 as a candidate) and neutron multiplier (Be) and rely on the development of advanced structural materials (a reduced activation ferritic steel F82H) compatible with high temperature operation. (author)

Conceptual design of Blanket Remote Handling System for CFETR

Energy Technology Data Exchange (ETDEWEB)

Wei, Jianghua, E-mail: weijh@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

2015-11-15

Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

Conceptual design of Blanket Remote Handling System for CFETR

International Nuclear Information System (INIS)

Wei, Jianghua; Song, Yuntao; Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong

2015-01-01

Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

Economic evaluation of the Blanket Comparison and Selection Study

International Nuclear Information System (INIS)

Waganer, L.M.

1985-01-01

The economic impact of employing the highly ranked blankets in the Blanket Comparison and Selection Study (BCSS) was evaluated in the context of both a tokamak and a tandem mirror power reactor (TMR). The economic evaluation criterion was determined to be the cost of electricity. The influencing factors that were considered are the direct cost of the blankets and related systems; the annual cost of blanket replacement; and the performance of the blanket, heat transfer, and energy conversion systems. The technical and cost bases for comparison were those of the STARFIRE and Mirror Advanced Reactor Study conceptual design power plants. The economic evaluation results indicated that the nitrate-salt-cooled blanket concept is an economically attractive concept for either reactor type. The water-cooled, solid breeder blanket is attractive for the tokamak and somewhat less attractive for the TMR. The helium-cooled, liquidlithium breeder blanket is the least economically desirable of higher ranked concepts. The remaining self-cooled liquid-metal and the helium-cooled blanket concepts represent moderately attractive concepts from an economic standpoint. These results are not in concert with those found in the other BCSS evaluation areas (engineering feasibility, safety, and research and development (R and D) requirements). The blankets faring well economically had generally lower cost components, lower pumping power requirements, and good power production capability. On the other hand, helium- and lithium-cooled systems were preferred from the standpoints of safety, engineering feasibility, and R and D requirements

Fission 2009 4. International Workshop on Nuclear Fission and Fission Product Spectroscopy - Compilation of slides

International Nuclear Information System (INIS)

2009-01-01

This conference is dedicated to the last achievements in experimental and theoretical aspects of the nuclear fission process. The topics include: mass, charge and energy distribution, dynamical aspect of the fission process, nuclear data evaluation, quasi-fission and fission lifetime in super heavy elements, fission fragment spectroscopy, cross-section and fission barrier, and neutron and gamma emission. This document gathers the program of the conference and the slides of the presentations

Nuclear fission and neutron-induced fission cross-sections

Energy Technology Data Exchange (ETDEWEB)

James, G.D.; Lynn, J.E.; Michaudon, A.; Rowlands, J.; de Saussure, G.

1981-01-01

A general presentation of current knowledge of the fission process is given with emphasis on the low energy fission of actinide nuclei and neutron induced fission. The need for and the required accuracy of fission cross section data in nuclear energy programs are discussed. A summary is given of the steps involved in fission cross section measurement and the range of available techniques. Methods of fission detection are described with emphasis on energy dependent changed and detector efficiency. Examples of cross section measurements are given and data reduction is discussed. The calculation of fission cross sections is discussed and relevant nuclear theory including the formation and decay of compound nuclei and energy level density is introduced. A description of a practical computation of fission cross sections is given.

Fusion blankets for high efficiency power cycles

International Nuclear Information System (INIS)

Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Usher, J.L.

1980-04-01

Definitions are given of 10 generic blanket types and the specific blanket chosen to be analyzed in detail from each of the 10 types. Dimensions, compositions, energy depositions and breeding ratios (where applicable) are presented for each of the 10 designs. Ultimately, based largely on neutronics and thermal hyraulics results, breeding an nonbreeding blanket options are selected for further design analysis and integration with a suitable power conversion subsystem

Simulations of fusion chamber dynamics and first wall response in a Z-pinch driven fusion–fission hybrid power reactor (Z-FFR)

Energy Technology Data Exchange (ETDEWEB)

Qi, J.M., E-mail: qjm06@sina.com [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China); Center for Fusion Energy Science and Technology (CFEST), China Academy of Engineering Physics, Mianyang 621999 (China); Wang, Z., E-mail: wangz_es@caep.cn [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China); Center for Fusion Energy Science and Technology (CFEST), China Academy of Engineering Physics, Mianyang 621999 (China); Chu, Y.Y., E-mail: chuyanyun@caep.cn [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China); Center for Fusion Energy Science and Technology (CFEST), China Academy of Engineering Physics, Mianyang 621999 (China); Li, Z.H., E-mail: lee_march@sina.com [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China)

2016-03-15

Highlights: • Z-FFR utilizes DT neutrons to drive a sub-critical fission blanket to produce energy. • A metal shell and Ar gas are employed in the fusion chamber for shock mitigation. • Massive materials can effectively mitigate the thermal heats on the chamber wall. • The W-coated Zr-alloy first wall exhibits good viability as a long-lived component. - Abstract: In a Z-pinch driven fusion–fission hybrid power reactor (Z-FFR), the fusion target will produce enormous energy of ∼1.5 GJ per pulse at a frequency of 0.1 Hz. Almost 20% of the fusion energy yield, approximately 300 MJ, is released in forms of pulsed X-rays. To prevent the first wall from fatal damages by the intense X-rays, a thin spherical metal shell and rare Ar buffer gas are introduced to mitigate the transient X-ray bursts. Radiation hydrodynamics in the fusion chamber were investigated by MULTI-1D simulations, and the corresponding thermal and mechanical loads on the first wall were also obtained. The simulations indicated that by optimizing the design parameters of the metal shell and Ar buffer gas, peak power flux of the thermal heats on the first wall could be mitigated to less than 10{sup 4} W/cm{sup 2} within a time scale of several milliseconds, while peak overpressures of the mechanical loads varying from 0.6 to 0.7 MPa. In addition, the thermomechanical response in a W–coated Zr-alloy first wall was performed by FWDR1D calculations using the derived thermal and mechanical loads as inputs. The temperature and stress fields were analyzed, and the corresponding elastic strains were conducted for primary lifetime estimations by using the Coffin–Manson relationships of both W and Zr-alloy. It was shown that the maximum temperature rises and stresses in the first wall were less than 50 K and 130 MPa respectively, and lifetime of the first wall would be in excess of 10{sup 9} cycles. The chamber exhibits good viability as a long-lived component to sustain the Z-FFR conceptual

Influence of surrounding environment on subcritical crack growth in marble

Science.gov (United States)

Nara, Yoshitaka; Kashiwaya, Koki; Nishida, Yuki; , Toshinori, Ii

2017-06-01

Understanding subcritical crack growth in rock is essential for determining appropriate measures to ensure the long-term integrity of rock masses surrounding structures and for construction from rock material. In this study, subcritical crack growth in marble was investigated experimentally, focusing on the influence of the surrounding environment on the relationship between the crack velocity and stress intensity factor. The crack velocity increased with increasing temperature and/or relative humidity. In all cases, the crack velocity increased with increasing stress intensity factor. However, for Carrara marble (CM) in air, we observed a region in which the crack velocity still increased with temperature, but the increase in the crack velocity with increasing stress intensity factor was not significant. This is similar to Region II of subcritical crack growth observed in glass in air. Region II in glass is controlled by mass transport to the crack tip. In the case of rock, the transport of water to the crack tip is important. In general, Region II is not observed for subcritical crack growth in rock materials, because rocks contain water. Because the porosity of CM is very low, the amount of water contained in the marble is also very small. Therefore, our results imply that we observed Region II in CM. Because the crack velocity increased in both water and air with increasing temperature and humidity, we concluded that dry conditions at low temperature are desirable for the long-term integrity of a carbonate rock mass. Additionally, mass transport to the crack tip is an important process for subcritical crack growth in rock with low porosity.

Electromagnetic analysis of ITER shield blanket under VDE

International Nuclear Information System (INIS)

Kang Weishan; Chen Jiming; Wu Jihong; Wang Mingxu

2010-01-01

Electromagnetic force and torque of ITER shield blanket system and their surrounding major component under vertical displacement event (VDE) were calculated with finite element method. ANSYS APDL was used to simulate the shape and magnitude of plasmas current dynamically in the VDE course, and external magnetic field was imposed, then the induced current distribution inside the all conductor including the blanket was obtained from the calculation. The force and torque for every blanket module was obtained to assess the safety of blanket system under VDE. (authors)

Neutronic design for the TFTR lithium blanket module

International Nuclear Information System (INIS)

Cheng, E.T.; Engholm, B.A.; Su, S.D.

1981-01-01

The preliminary design of a lithium blanket module (LBM) to be installed and tested in the TFTR has been performed under subcontract to PPPL and EPRI. The objectives of the LBM program are calculation and measurement of neutron fluences and tritium production in a breeding blanket module using state of art techniques, comparison of calculations with measurements, and acquisition of operational experience with a fusion reactor blanket module. The neutronic design of the LBM is one of the key areas of this program in which the LBM composition and geometry are optimized and the boundary material effects on the tritium production in the blanket module are explored. The concept of employing sintered Li/sub 2/O pellets in tubes is proposed for the blanket design

LMFBR Blanket Physics Project progress report No. 2

International Nuclear Information System (INIS)

Forbes, I.A.; Driscoll, M.J.; Rasmussen, N.C.; Lanning, D.D.; Kaplan, I.

1971-01-01

This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238 U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics

Burning of spent fuel of an accelerator-driven modular HTGR in sub-critical condition

International Nuclear Information System (INIS)

Jing Xingqing; Yang Yongwei; Chang Hong; Wu Zongxin; Gu Yuxiang

2002-01-01

The modular high temperature gas cooled reactor (MHTGR) has good safety characteristics because of the use of coated particles in the fuel element. After the particles cool outside of the reactor for some time, the spent fuel can be re-utilized. The author describes a physics feasibility study for the burning of spent fuel from a 350 MW ring-shaped modular high temperature gas cooled reactor in an accelerator-driven sub-critical reactor. A conceptual design is given for the 30 MW accelerator-driven sub-critical reactor. The neutron transport in the sub-critical reactor was simulated using the MCNP code, and the burnup was calculated using the ORIGEN2 code. The results show that the accelerator-driven sub-critical gas-cooled reactor has reliable sub-criticality and low power density and that the spent fuel from a 350 MW ring-shaped modular high temperature gas cooled reactor can be burned to provide 20% more energy

Development of High Flux Isotope Reactor (HFIR) subcriticality monitoring methods

International Nuclear Information System (INIS)

Rothrock, R.B.

1991-01-01

Use of subcritical source multiplication measurements during refueling has been investigated as a possible replacement for out-of-reactor subcriticality measurements formerly made on fresh HFIR fuel elements at the ORNL Critical Experiment Facility. These measurements have been used in the past for preparation of estimated critical rod positions, and as a partial verification, prior to reactor startup, that the requirements for operational shutdown margin would be met. Results of subcritical count rate data collection during recent HFIR refuelings and supporting calculations are described illustrating the intended measurement method and its expected uncertainty. These results are compared to historical uses of the out-of-reactor core measurements and their accuracy requirements, and a planned in-reactor test is described which will establish the sensitivity of the method and calibrate it for future routine use during HFIR refueling. 2 refs., 1 fig., 2 tabs

Thorium utilization in a small long-life HTR. Part II: Seed-and-blanket fuel blocks

Energy Technology Data Exchange (ETDEWEB)

Ding, Ming, E-mail: dingming@hrbeu.edu.cn [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

2014-02-15

Highlights: • Seed-and-blanket (S and B) fuel blocks are proposed for a small block-type HTR. • S and B fuel blocks consist of a seed region (UO{sub 2}) and a blanket region (ThO{sub 2}). • The neutronic performance of S and B fuel blocks are analyzed using SCALE 6. • Three S and B fuel blocks with a reactivity swing of 0.1 Δk are recommended. • S and B fuel blocks are compared with thorium MOX fuel blocks. - Abstract: In order to utilize thorium in high temperature gas-cooled reactors (HTRs), the concept of seed-and-blanket (S and B) fuel block is introduced into the U-Battery, which is a long-life block-type HTR with a thermal power of 20 MWth. A S and B fuel block consists of a seed region with uranium in the center, and a blanket region with thorium. The neutronic performance, such as the multiplication factor, conversion ratio and reactivity swing, of a typical S and B fuel block was investigated by SCALE 6.0 by parametric analysis of the composition parameters and geometric parameters of the fuel block for the U-Battery application. Since the purpose of U-235 in the S and B fuel block is to ignite the fission reactions in the fuel block, 20% enriched uranium is recommended for the S and B fuel block. When the ratio of the number of carbon to heavy metal atoms changes with the geometric parameters of the fuel block in the range of 200–250, the reactivity swing reaches very small values. Furthermore, for a reactivity swing of 0.1 Δk during 10 effective full power years, three configurations with 36, 54 and 78 UO{sub 2} fuel rods are recommended for the application of the U-Battery. The comparison analysis of the S and B fuel block with the Th/U MOX fuel block shows that the former has a longer lifetime and a lower reactivity swing.

Advanced high performance solid wall blanket concepts

International Nuclear Information System (INIS)

Wong, C.P.C.; Malang, S.; Nishio, S.; Raffray, R.; Sagara, A.

2002-01-01

First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability

Nuclear fission and fission-product spectroscopy: 3. International workshop on nuclear fission and fission-product spectroscopy

International Nuclear Information System (INIS)

Goutte, Heloise; Fioni, Gabriele; Faust, Herbert; Goutte, Dominique

2005-01-01

The present book contains the proceedings of the third workshop in a series of workshops previously held in Seyssins in 1994 and 1998. The meeting was jointly organized by different divisions of CEA and two major international laboratories. In the opening address, Prof. B. Bigot, the French High Commissioner for Atomic Energy, outlined France's energy policy for the next few decades. He emphasized the continuing progress of nuclear fission in both technical and economic terms, allowing it to contribute to the energy needs of the planet even more in the future than it does today. Such progress implies a very strong link between fundamental and applied research based on experimental and theoretical approaches. The workshop gathered the different nuclear communities studying the fission process, including topics as the following: - nuclear fission experiments, - spectroscopy of neutron rich nuclei, - fission data evaluation, - theoretical aspects of nuclear fission, - and innovative nuclear systems and new facilities. The scientific program was suggested by an International Advisory Committee. About 100 scientists from 13 different countries attended the conference in the friendly working atmosphere of the Castle of Cadarache in the heart of the Provence. The proceedings of the workshop were divided into 11 sections addressing the following subject matters: 1. Cross sections and resonances (5 papers); 2. Fission at higher energies - I (5 papers); 3. Fission: mass and charge yields (4 papers); 4. Light particles and cluster emission (4 papers); 5. Spectroscopy of neutron rich nuclei (5 papers); 6. Resonances, barriers, and fission times (5 papers); 7. Fragment excitation and neutron emission (4 papers); 8. Mass and energy distributions (4 papers); 9. Needs for nuclear data and new facilities - I (4 papers); 10. Angular momenta and fission at higher Energies - II (3 papers); 11. New facilities - II (2 papers). A poster session of 8 presentations completed the workshop

Estimation of subcriticality with the computed values. 2

Energy Technology Data Exchange (ETDEWEB)

Sakurai, Kiyoshi; Arakawa, Takuya; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1996-02-01

For measurements of reactivities and neutron count rate space distributions, seven subcritical cores including non-square array cores were constructed using critical assembly TCA. MCNP-4A was used for the experimental analysis. The calculational results of the neutron count rate space distributions agreed with the measured ones within the each error range. It means that for calculation error indirect estimation method, the calculated neutron multiplication factors are equal to ones of experimental reactivities. It is shown that from these experiments and calculations estimation method of subcriticality with the computed values based on the calculation error indirect estimation method is also applicable to six non-square array cores. (author).

Liquid metal cooled blanket concept for NET

International Nuclear Information System (INIS)

Malang, S.; Casal, V.; Arheidt, K.; Fischer, U.; Link, W.; Rust, K.

1986-01-01

A blanket concept for NET using liquid lithium-lead both as breeder material and as coolant is described. The need for inboard breeding is avoided by using beryllium as neutron multiplier in the outboard blanket. Novel flow channel inserts are employed in all poloidal ducts to reduce the MHD pressure drop. The concept offers a simple mechanical design and a higher tritium breeding ratio compared to water- and gas-cooled blankets. (author)

Beryllium R&D for blanket application

Science.gov (United States)

Donne, M. Dalle; Longhurst, G. R.; Kawamura, H.; Scaffidi-Argentina, F.

1998-10-01

The paper describes the main problems and the R&D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point.

Heat transfer problems in gas-cooled solid blankets

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1976-01-01

In all fusion reactors using the deuterium-tritium fuel cycle, a large fraction approximately 80 percent of the fusion energy will be released as approximately 14 MeV neutrons which must be slowed down in a relatively thick blanket surrounding the plasma, thereby, converting their kinetic energy to high temperature heat which can be continuously removed by a coolant stream and converted in part to electricity in a conventional power turbine. Because of the primary goal of achieving minimum radioactivity, to date Brookhaven blanket concepts have been restricted to the use of some form of solid lithium, with inert gas-cooling and in some design cases, water-cooling of the shell structure. Aluminum and graphite have been identified as very promising structural materials for fusion blankets, and conceptual designs based on these materials have been made. Depending on the thermal loading on the ''first'' wall which surrounds the plasma as well as blanket design, heat transfer problems may be noticeably different in gas-cooled solid blankets. Approaches to solution of heat removal problems as well as explanation of: (a) the after-heat problems in blankets; (b) tritium breeding in solids; and (c) materials selection for radiation shields relative to the minimum activity blanket efforts at Brookhaven are discussed

Accelerator-driven assembly for plutonium transformation (ADAPT)

Science.gov (United States)

Tuyle, Greorgy J. Van; Todosow, Michael; Powell, James; Schweitzer, Donald

1995-01-01

A particle accelerator-driven spallation target and corresponding blanket region are proposed for the ultimate disposition of weapons-grade plutonium being retired from excess nuclear weapons in the U.S. and Russia. The highly fissle plutonium is contained within .25 to .5 cm diameter silicon-carbide coated graphite beads, which are cooled by helium, within the slightly subcritical blanket region. Major advantages include very high one-pass burnup (over 90%), a high integrity waste form (the coated beads), and operation in a subcritical mode, thereby minimizing the vulnerability to the positive reativity feedbacks often associated with plutonium fuel.

Blanket design study for a Commercial Tokamak Hybrid Reactor (CTHR)

International Nuclear Information System (INIS)

Chapin, D.L.; Green, L.; Lee, A.Y.; Culbert, M.E.; Kelly, J.L.

1979-09-01

The results are presented of a study on two blanket design concepts for application in a Commercial Tokamak Hybrid Reactor (CTHR). Both blankets operate on the U-Pu cycle and are designed to achieve tritium self-sufficiency while maximizing the fissile fuel production within thermal and mechanical design constraints. The two blanket concepts that were evaluated were: (1) a UC fueled, stainless steel clad and structure, helium cooled blanket; and (2) a UO 2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li 2 O and LiH, were evaluated for use in both blanket concepts. The use of lead as a neutron multiplier or reflector and graphite as a reflector was also considered for both blankets

Reducing beryllium content in mixed bed solid-type breeder blankets

Energy Technology Data Exchange (ETDEWEB)

Shimwell, J., E-mail: mail@jshimwell.com [Department of Physics and Astronomy, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom); Lilley, S.; Morgan, L.; Packer, L.; Kovari, M.; Zheng, S. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); McMillan, J. [Department of Physics and Astronomy, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom)

2016-11-01

Highlights: • The ratio of breeder ceramic to neutron multiplier of breeder blankets was varied linearly with depth. • Blankets with varying composition were found to perform better than uniform composition breeder blankets. • It was also possible to reduce the amount of beryllium required by the blanket. - Abstract: Beryllium (Be) is a precious resource with many high value uses, the low energy threshold (n,2n) reaction makes Be an excellent neutron multiplier for use in fusion breeder blankets. Estimates of Be requirements and available resources suggest that this could represent a major supply difficulty for solid-type blanket concepts. Reducing the quantity of Be required by breeder blankets would help to alleviate the problem to some extent. In addition, it is important that the reduction in the Be quantity does not diminish the blanket's performance in key aspects such as the tritium breeding ratio (TBR), energy multiplication and peak nuclear heating. Mixed pebble bed designs allow for the multiplier fraction to be varied throughout the blanket. This neutronics study used MCNP 6 to investigate linear variations of the multiplier fraction in relation to blanket depth, in order to better utilise the important multiplying Be(n,2n) and breeding reactions. Blankets with a uniform multiplier fraction showed little scope for reduction in Be mass. Blankets with varying multiplier fractions were able to simultaneously use 10% less Be, increase the energy amplification by 1%, reduce the peak heating by 7% and maintaining a sufficient TBR when compared to the performance achievable using a uniform composition.

Thermohydraulics design and thermomechanics analysis of two European breeder blanket concepts for DEMO. Pt. 1 and Pt. 2. Pt. 1: BOT helium cooled solid breeding blanket. Pt. 2: Dual coolant self-cooled liquid metal blanket

International Nuclear Information System (INIS)

Norajitra, P.

1995-06-01

Two different breeding blanket concepts are being elaborated at Forschungszentrum Karlsruhe within the framework of the DEMO breeding blanket development, the concept of a helium cooled solid breeding blanket and the concept of a self-cooled liquid metal blanket. The breeder material used in the first concept is Li 4 SiO 4 as a pebble bed arranged separate from the beryllium pebble bed, which serves as multiplier. The breeder material zone is cooled by several toroidally-radially configurated helium cooling plates which, at the same time, act as reinforcements of the blanket structures. In the liquid metal blanket concept lead-lithium is used both as the breeder material and the coolant. It flows at low velocity in poloidal direction downwards and back in the blanket front zone. In both concepts the First Wall is cooled by helium gas. This report deals with the thermohydraulics design and thermomechanics analysis of the two blanket concepts. The performance data derived from the Monte-Carlo computations serve as a basis for the design calculations. The coolant inlet and outlet temperatures are chosen with the design criteria and the economics aspects taken into account. Uniform temperature distribution in the blanket structures can be achieved by suitable branching and routing of the coolant flows which contributes to reducing decisively the thermal stress. The computations were made using the ABAQUS computer code. The results obtained of the stresses have been evaluated using the ASME code. It can be demonstrated that all maximum values of temperature and stress are below the admissible limit. (orig.) [de

Experimental determination of the neutron source for the Argonauta reactor subcritical assembly

Energy Technology Data Exchange (ETDEWEB)

Renke, Carlos A.C.; Furieri, Rosanne C.A.A.; Pereira, Joao C.S.; Voi, Dante L.; Barbosa, Andre L.N., E-mail: renke@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2011-07-01

The utilization of a subcritical assembly for the determination of nuclear parameters in a multiplier medium requires a well defined neutron source to carry out the experiments necessary for the acquisition of the desired data. The Argonauta research reactor installed at the Instituto de Engenharia Nuclear has a subcritical assembly, under development, to be coupled at the upper part of the reactor core that will provide the needed neutrons emerging from its internal thermal column made of graphite. In order to perform neutronic calculations to compare with the experimental results, it is necessary a precise knowledge of the emergent neutron flux that will be used as neutron source in the subcritical assembly. In this work, we present the thermal neutron flux profile determined experimentally via the technique of neutron activation analysis, using dysprosium wires uniformly distributed at the top of the internal thermal neutron column of the Argonauta reactor and later submitted to a detection system using Geiger-Mueller detector. These experimental data were then compared with those obtained through neutronic calculation using HAMMER and CITATION codes in order to validate this calculation system and to define a correct neutron source distribution to be used in the subcritical assembly. This procedure avoids a coupled neutronic calculation of the subcritical assembly and the reactor core. It has also been determined the dimension of the graphite pedestal to be used in the bottom of the subcritical assembly tank in order to smooth the emergent neutron flux at the reactor top. Finally, it is estimated the thermal neutron flux inside the assembly tank when filled with water. (author)

MIT LMFBR blanket research project. Final summary report

International Nuclear Information System (INIS)

Driscoll, M.J.

1983-08-01

This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record

Delayed fission

Energy Technology Data Exchange (ETDEWEB)

Hatsukawa, Yuichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-07-01

Delayed fission is a nuclear decay process that couples {beta} decay and fission. In the delayed fission process, a parent nucleus undergoes {beta} decay and thereby populates excited states in the daughter. If these states are of energies comparable to or greater than the fission barrier of the daughter, then fission may compete with other decay modes of the excited states in the daughter. In this paper, mechanism and some experiments of the delayed fission will be discussed. (author)

Thermomechanical analysis of the DFLL test blanket module for ITER

International Nuclear Information System (INIS)

Chen Hongli; Wu Yican; Bai Yunqing

2006-01-01

The finite element code is used to simulate two kinds of blanket design structure, which are SLL (Quasi-Static Lithium Lead) and DLL (Dual-cooled Lithium Lead) blanket concepts for the Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM) submitted to the ITER test blanket working group. The temperature and stress distributions have been presented for the two kinds of blanket structure on the basis of the structural design, thermal-hydraulic design and neutronics analysis. Also the mechanical performance is presented for the high temperature component of blanket structure according to the ITER Structural Design Criteria (ISDC). The rationality and feasibility of the two kinds of blanket structure design of DFLL-TBM have been analyzed based on the above results which also acted as the theoretical base for further optimized analysis. (authors)

Accelerator-driven system design concept for disposing of spent nuclear fuels

International Nuclear Information System (INIS)

Gohar, Y.; Cao, Y.; Kellogg, R.; Merzari, E.

2015-01-01

At present, the US SNF (Spent Nuclear Fuel) inventory is growing by about 2,000 metric tonnes (MT) per year from the current operating nuclear power plants to reach about 70,000 MT by 2015. This SNF inventory contains about 1% transuranics (700 MT), which has about 115 MT of minor actinides. Accelerator-driven systems utilising proton accelerators with neutron spallation targets and subcritical blankets can be utilised for transmuting these transuranics, simultaneously generating carbon free energy, and significantly reducing the capacity of the required geological repository storage facility for the spent nuclear fuels. A fraction of the SNF plutonium can be used as a MOX fuel in the current/future thermal power reactors and as a starting fuel for future fast power reactors. The uranium of the spent nuclear fuel can be recycled for use in future nuclear power plants. This paper shows that only four to five accelerator-driven systems operating for less than 33 full power years can dispose of the US SNF inventory expected by 2015. In addition, a significant fraction of the long-lived fission products will be transmuted at the same time. Each system consists of a proton accelerator with a neutron spallation target and a subcritical assembly. The accelerator beam parameters are 1 GeV protons and 25 MW beam power, which produce 3 GWt in the subcritical assembly. A liquid metal (lead or lead-bismuth eutectic) spallation target is selected because of design advantages. This target is located at the centre of the subcritical assembly to maximise the utilisation of spallation neutrons. Because of the high power density in the target material, the target has its own coolant loop, which is independent of the subcritical assembly coolant loop. Mobile fuel forms with transuranic materials without uranium are considered in this work with liquid lead or lead-bismuth eutectic as fuel carrier

Basic principles of lead and lead-bismuth eutectic application in blanket of fusion reactors

International Nuclear Information System (INIS)

Beznosov, A.V.; Pinaev, S.S.; Muraviev, E.V.; Romanov, P.V.

2005-01-01

High magnetohydrodynamic pressure drop is an important issue for liquid metal blanket concepts. To decrease magnetohydrodynamic resistance authors propose to form insulating coatings on internal surface of blanket ducts at any moment of fusion reactor exploitation. It may be achieved easily if lead or lead-bismuth eutectic is used and technology of oxidative potential handling is applied. A number of experiments carried out in NNSTU show the availability of the proposed technology. It bases on formation of the insulating coatings that consist of the oxides of components of the structural materials and of the coolant components. In-situ value of the insulating coatings characteristics ρδ is ∼ 10 -5 Ohm·m 2 for steels and 5,0x10 -6 - 5,0x10 -5 Ohm·m 2 for vanadium alloys. Thermal cycling is possible during exploitation of a blanket. The experimental research of the insulating coatings properties during thermal cycling have shown that the coatings formed into the lead and lead-bismuth coolants save there insulating properties. Experience of many years is an undoubted advantage of the lead-bismuth coolant and less of the lead coolant in comparison with lithium. Russian Federation possesses of experience of exploitation of the research and industrial facilities, of experience of creation of the pumps, steamgenerators and equipment with heavy liquid metal coolants. The unique experience of designing, assembling and exploitation of the fission reactors with lead-bismuth coolant is also available. The problem of technology of lead and lead-bismuth coolants for power high temperature radioactive facilities has been solved. Accidents, emergency situations such as leakage of steamgenerators or depressurization of gas system in facilities with lead and lead-bismuth coolants have been explored and suppressed. (author)

Parametric systems analysis for ICF hybrid reactors

International Nuclear Information System (INIS)

Berwald, D.H.; Maniscalco, J.A.; Chapin, D.L.

1981-01-01

Parametric design and systems analysis for inertial confinement fusion-fission hybrids are presented. These results were generated as part of the Electric Power Research Institute (EPRI) sponsored Feasibility Assessment of Fusion-Fission Hybrids, using an Inertial Confinement Fusion (ICF) hybrid power plant design code developed in conjunction with the feasibility assessment. The SYMECON systems analysis code, developed by Westinghouse, was used to generate economic results for symbiotic electricity generation systems consisting of the hybrid and its client Light Water Reactors (LWRs). These results explore the entire fusion parameter space for uranium fast fission blanket hybrids, thorium fast fission blanket hybrids, and thorium suppressed fission blanket types are discussed, and system sensitivities to design uncertainties are explored

Blanket options for high-efficiency fusion power

International Nuclear Information System (INIS)

Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

1980-01-01

The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

Fusion blankets for high-efficiency power cycles

International Nuclear Information System (INIS)

Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

1980-01-01

The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

Fusion blanket for high-efficiency power cycles

International Nuclear Information System (INIS)

Usher, J.L.; Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Taussig, R.

1980-01-01

The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperature (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by Ar) utilizing Li 2 O for tritium breeding. In this design, approx. 60% of the fusion energy is deposited in the high-temperature interior. The maximum Ar temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

Fusion blankets for high-efficiency power cycles

International Nuclear Information System (INIS)

Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

1981-01-01

The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 deg C) of conventional structural materials such as stainless steels. In this project 'two-zone' blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 deg C leading to an overall efficiency estimate of 55 to 60% for this reference case. (author)

Investigation of (n, 2n) reaction and fission rates in iron-shielded uranium samples bombarded by 14.9 MeV neutrons

International Nuclear Information System (INIS)

Shani, G.

1976-01-01

The effect of the thickness of iron shielding on the (n, 2n) reaction rate in a fusion reactor (hybrid) blanket is investigated. The results are compared with the fission rate-dependence. Samples of natural uranium are irradiated with 14 MeV neutrons, with iron slabs of various thickness between the neutron generator target and the samples. Both reactions are threshold reactions but the fact that the 238 U (n, 2n) reaction threshold is at 6 MeV and that of fission is at 2 MeV makes the ratio between the two very much geometry-dependent. Two geometrical effects take place, the 1/r 2 and the build-up. While the build-up affects the (n, 2n) reaction rate, the fission rate is affected more by the 1/r 2 effect. The reason is that both elastic and inelastic scattering end up with neutrons with energy above fission threshold, while only elastic scattering brings high energy neutrons to the sample and causes (n, 2n) reaction. A comparison is made with calculated results where the geometrical effects do not exist. (author)

Solubility and degradation of paracetamol in subcritical water

Directory of Open Access Journals (Sweden)

Emire Zuhal

2017-01-01

Full Text Available In this study, solubility and degradation of paracetamol were examined using subcritical water. Effect of temperature and static time was investigated during solubility process in subcritical water at constant pressure (50 bar. Experimental results show that temperature and static time have crucial effect on the degradation and solubility rates. Maximum mole fraction for solubility of paracetamol was obtained at 403 K as (14.68 ± 0.74×103. Approximation model for solubility of paracetamol was proposed. O2 and H2O2 were used in degradation process of paracetamol. Maximum degradation rate was found as 68.66 ± 1.05 and 100 ± 0.00 % using O2 and H2O2, respectively.

Experimental subcritical facility driven by D-D/D-T neutron generator at BARC, India

Energy Technology Data Exchange (ETDEWEB)

Sinha, Amar, E-mail: image@barc.gov.in; Roy, Tushar; Kashyap, Yogesh; Ray, Nirmal; Shukla, Mayank; Patel, Tarun; Bajpai, Shefali; Sarkar, P.S.; Bishnoi, Saroj

2015-05-01

Highlights: •Experimental subcritical facility BRAHMMA coupled to D-D/D-T neutron generator. •Preliminary results of PNS experiments reported. •Feynman-alpha noise measurements explored with continuous source. -- Abstract: The paper presents design of an experimental subcritical assembly driven by D-D/D-T neutron and preliminary experimental measurements. The system has been developed for investigating the static and dynamic neutronic properties of accelerator driven sub-critical systems. This system is modular in design and it is first in the series of subcritical assemblies being designed. The subcritical core consists of natural uranium fuel with high density polyethylene as moderator and beryllium oxide as reflector. The fuel is embedded in high density polyethylene moderator matrix. Estimated k{sub eff} of the system is ∼0.89. One of the unique features of subcritical core is the use of Beryllium oxide (BeO) as reflector and HDPE as moderator making the assembly a compact modular system. The subcritical core is coupled to Purnima Neutron Generator which works in D-D and D-T mode with both DC and pulsed operation. It has facility for online source strength monitoring using neutron tagging and programmable source modulation. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques with D-D neutrons. Further experiments are being planned to measure the reactivity and other kinetic parameters using noise methods. This facility would also be used for carrying out studies on effect of source importance and measurement of source multiplication factor k{sub s} and external neutron source efficiency φ{sup ∗} in great details. Experiments with D-T neutrons are also underway.

Methods to enhance blanket power density

International Nuclear Information System (INIS)

Hsu, P.Y.; Miller, L.G.; Bohn, T.S.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Wessol, D.E.; Abdou, M.A.

1982-06-01

The overall objective of this task is to investigate the extent to which the power density in the FED/INTOR breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow advanced reactor blanket modules to be tested on FED/INTOR under representative conditions

Neutronic optimization of solid breeder blankets for STARFIRE design

International Nuclear Information System (INIS)

Gohar, Y.; Abdou, M.A.

1980-01-01

Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture

Development of advanced blanket materials for solid breeder blanket of fusion reactor

International Nuclear Information System (INIS)

Ishitsuka, E.

2002-01-01

Advanced solid breeding blanket design in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high dose of neutron irradiation. Therefore, the development of such advanced blanket materials is indispensable. In this paper, the cooperation activities among JAERI, universities and industries in Japan on the development of these advanced materials are reported. Advanced tritium breeding material to prevent the grain growth in high temperature had to be developed because the tritium release behavior degraded by the grain growth. As one of such materials, TiO 2 -doped Li 2 TiO 3 has been studied, and TiO 2 -doped Li 2 TiO 3 pebbles was successfully fabricated. For the advanced neutron multiplier, the beryllium intermetallic compounds that have high melting point and good chemical stability have been studied. Some characterization of Be 12 Ti was studied. The pebble fabrication study for Be 12 Ti was also performed and Be 12 Ti pebbles were successfully fabricated. From these activities, the bright prospect to realize the DEMO blanket by the application of TiO 2 -doped Li 2 TiO 3 and beryllium intermetallic compounds was obtained. (author)

An evaluation of neutron and gamme heating in fission product isotopes

International Nuclear Information System (INIS)

Leal, L.C.; Hill, R.N.; Khalil, H.S.

1993-01-01

The accurate prediction of the energy deposition rate in fast reactors, particularly in blanket and nonfueled regions, requires explicit treatment of gamma photon transport. Such an explicit treatment is part of the coupled neutron-photon heating method in use at Argonne National Laboratory, (ANL). In applying this procedure, three approximations are made in connection with the modeling of fission products (FPs): 1. The contribution of the FP neutron interactions to the gamma source is neglected. 2. In computing the macroscopic gamma interaction cross sections, the FPs are either neglected or simulated with an element (usually molybdenum) representative of an open-quotes averageclose quotes FP. 3. The heating contribution of the FP is neglected by use of zero FP kerma factors

Computational investigation of 99Mo, 89Sr, and 131I production rates in a subcritical UO2(NO32 aqueous solution reactor driven by a 30-MeV proton accelerator

Directory of Open Access Journals (Sweden)

Z. Gholamzadeh

2015-12-01

Full Text Available The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing 99Mo. In this method, the medical isotope production system itself is used to extract 99Mo or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of 99Mo by irradiating targets. In this study, the neutronic performance and 99Mo, 89Sr, and 131I production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ∼1,500 Ci/wk (∼325 6-day Ci of 99Mo at the end of a cycle.

Economic analysis of fusion breeders. Supplement

International Nuclear Information System (INIS)

Delene, J.G.

1985-01-01

Three fusion/fission hybrids and three converter reactors were considered in combination: (1) Li-Be (Opt-Li) blanket, (2) molten salt blanket (1.6 blanket energy multiplier), and (3) molten salt blanket (2.5 blanket energy multiplier). The following converter (fission) reactors were considered: (1) LWR, (2) HTGR, and (3) molten salt. In order to provide some perspective on the results of the hybrid analysis, LMFBRs were also examined: (1) methods applied consistently, and (2) range of LMFBR costs consistent with current thought on advanced designs

Accelerator-driven subcritical facility:Conceptual design development

Science.gov (United States)

Gohar, Yousry; Bolshinsky, Igor; Naberezhnev, Dmitry; Duo, Jose; Belch, Henry; Bailey, James

2006-06-01

A conceptual design development of an accelerator-driven subcritical facility has been carried out in the preparation of a joint activity with Kharkov Institute of Physics and Technology of Ukraine. The main functions of the facility are the medical isotope production and the support of the Ukraine nuclear industry. An electron accelerator is considered to drive the subcritical assembly. The neutron source intensity and spectrum have been studied. The energy deposition, spatial neutron generation, neutron utilization fraction, and target dimensions have been quantified to define the main target performance parameters, and to select the target material and beam parameters. Different target conceptual designs have been developed based the engineering requirements including heat transfer, thermal hydraulics, structure, and material issues. The subcritical assembly is designed to obtain the highest possible neutron flux level with a Keff of 0.98. Different fuel materials, uranium enrichments, and reflector materials are considered in the design process. The possibility of using low enrichment uranium without penalizing the facility performance is carefully evaluated. The mechanical design of the facility has been developed to maximize its utility and minimize the time for replacing the target and the fuel assemblies. Safety, reliability, and environmental considerations are included in the facility conceptual design. The facility is configured to accommodate future design improvements, upgrades, and new missions. In addition, it has large design margins to accommodate different operating conditions and parameters. In this paper, the conceptual design and the design analyses of the facility will be presented.

Monte Carlo modeling and analyses of YALINA- booster subcritical assembly Part II: pulsed neutron source

International Nuclear Information System (INIS)

Talamo, A.; Gohar, M.Y.A.; Rabiti, C.

2008-01-01

One of the most reliable experimental methods for measuring the kinetic parameters of a subcritical assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology for characterizing the kinetic parameters of a subcritical assembly using the Sjoestrand method, which allows comparing the analytical and experimental time dependent reaction rates and the reactivity measurements. In this methodology, the reaction rate, detector response, is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the fission delayed neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction is vanished. The obtained reaction rate is superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The new calculation methodology has shown an excellent agreement with the experimental results available from the YALINA-Booster facility of Belarus. The facility has been driven by a Deuterium-Deuterium or Deuterium-Tritium pulsed neutron source and the (n,p) reaction rate has been experimentally measured by a 3 He detector. The MCNP calculation has utilized the weight window and delayed neutron biasing variance reduction techniques since the detector volume is small compared to the assembly volume. Finally, this methodology was used to calculate the IAEA benchmark of the YALINA-Booster experiment

Review: BNL Tokamak graphite blanket design concepts

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1976-01-01

The BNL minimum activity graphite blanket designs are reviewed, and three are discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a 30 cm or thicker graphite screen. Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy, which is then radiated to a secondary blanket with coolant tubes, as in types A and B, or removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma. (Auth.)

LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

Energy Technology Data Exchange (ETDEWEB)

Farmer, J

2008-09-08

The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

Environmental considerations for alternative fusion reactor blankets

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Young, J.R.

1975-01-01

Comparisons of alternative fusion reactor blanket/coolant systems suggest that environmental considerations will enter strongly into selection of design and materials. Liquid blankets and coolants tend to maximize transport of radioactive corrosion products. Liquid lithium interacts strongly with tritium, minimizing permeation and escape of gaseous tritium in accidents. However, liquid lithium coolants tend to create large tritium inventories and have a large fire potential compared to flibe and solid blankets. Helium coolants minimize radiation transport, but do not have ability to bind the tritium in case of accidental releases. (auth)

A Li-particulate blanket concept for ITER

International Nuclear Information System (INIS)

Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.

1989-01-01

The Li-particulate blanket design concept the authors proposed for the International Thermonuclear Experimental Reactor (ITER) uses a dilute suspension of fine solid breeder particles in a carrier gas as the combined coolant and lithium breeder stream. This blanket concept has a simple mechanical and hydraulic configuration, low inventory of bred tritium, and simple tritium extraction system. Existing technology can be used to implement the design for ITER. The concept has the potential to be a highly reliable shield and blanket design for ITER with relatively low development and capital costs

Development study on subcriticality monitor. 1. Report under business contract with Japan Nuclear Fuel Cycle Development Institute

CERN Document Server

Yamada, S

2002-01-01

In this trust fund, we reviewed subcriticality measuring methods and neutron or gamma ray measuring and date transmission systems appropriate for realizing inexpensive on-line criticality surveillance systems, which is required for ensuring the safety of nuclear fuel reprocessing plants. Since the neutron flux level in subcritical systems is fairly low without external neutron sources, it is desirable to use pulse type neutron detectors for subcritical measurement systems. This logically implies that subcriticality measurement methods based on the temporal domain should be used for developing an on-line criticality surveillance system. In the deep subcriticality conditions, a strong external neutron source is needed for eactivity measurement and a D-T tube can be used in order to improve the accuracy of the measurement. A D-T tube is convenient since it is free from Tritium problem since Tritium is sealed in an airtight container and also can be controlled by power supply. Hence, under deep subcritical condit...

Coulomb fission and transfer fission at heavy ion collisions

International Nuclear Information System (INIS)

Himmele, G.

1981-01-01

In the present thesis the first direct evidence of nuclear fission after inelastic scattering of heavy ions (sup(183,184)W, 152 Sm → 238 U; 184 W → 232 Th; 184 W, 232 Th → 248 Cm) is reported. Experiments which were performed at the UNILAC of the Gesellschaft fuer Schwerionenforschung in Darmstadt show the observed heavy ion induced fission possesses significant properties of the Coulomb fission. The observed dependence of the fission probability for inelastic scattering on the projectile charge proves that the nuclear fission is mediated by the electromagnetic interaction between heavy ions. This result suggests moreover a multiple Coulomb-excitation preceding the fission. Model calculations give a first indication, that the Coulomb fission proceeds mainly from the higher β phonons. In the irradiation with 184 W the fission probability of 232 Th is for all incident energies about 40% smaller that at 238 U. The target dependence of the Coulomb fission however doesn't allow, to give quantitative statements about the position and B(E2)-values of higher lying β phonons. (orig./HSI) [de

Designing a mini subcritical nuclear reactor; Diseno de un mini reactor nuclear subcritico

Energy Technology Data Exchange (ETDEWEB)

Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M., E-mail: rafelaescobedo@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Jardin Juarez 147, Col. Centro, 98000 Zacatecas, Zac. (Mexico)

2015-10-15

In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of {sup 239}PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of {sup 239}PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k{sub e}-f{sub f}, the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k{sub eff}, the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons {sup 239}PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k{sub eff} was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k{sub eff} of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five

Neutron dosimetry for the TFTR Lithium-Blanket-Module program

International Nuclear Information System (INIS)

Harker, Y.D.; Tsang, F.Y.; Caffrey, A.J.; Homeyer, W.G.; Engholm, B.A.

1981-01-01

The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-a-kind neutronics experiment involving a prototypical fusion reactor blanket module with a distributed neutron source from the plasma of the TFTR at Princeton Plasma Physics Laboratory. The objectives of the LBM program are: (1) to test the capabilities of neutron transport codes when applied to fusion test reactor blanket conditions, and (2) to obtain tritium breeding performance data on a typical design concept of a fusion-reactor blanket. This paper addresses the issues relative to the measurement of neutron fields in the LBM, presents the results of preliminary design studies concerning neutron measurements and also presents the results of blanket mockup experiments performed at the Idaho National Engineering Laboratory

Packed-fluidized-bed blanket concept for a thorium-fueled commercial tokamak hybrid reactor

International Nuclear Information System (INIS)

Chi, J.W.H.; Miller, J.W.; Karbowski, J.S.; Chapin, D.L.; Kelly, J.L.

1980-09-01

A preliminary design of a thorium blanket was carried out as a part of the Commercial Tokamak Hybrid Reactor (CTHR) study. A fixed fuel blanket concept was developed as the reference CTHR blanket with uranium carbide fuel and helium coolant. A fixed fuel blanket was initially evaluated for the thorium blanket study. Subsequently, a new type of hybrid blanket, a packed-fluidized bed (PFB), was conceived. The PFB blanket concept has a number of unique features that may solve some of the problems encountered in the design of tokamak hybrid reactor blankets. This report documents the thorium blanket study and describes the feasibility assessment of the PFB blanket concept

A Feasible DEMO Blanket Concept Based on Water Cooled Solid Breeder

Energy Technology Data Exchange (ETDEWEB)

Someya, Y.; Tobita, K.; Utoh, H.; Hoshino, K.; Asakura, N.; Nakamura, M.; Tanigawa, H.; Mikio, E.; Tanigawa, H.; Nakamichi, M.; Hoshino, T., E-mail: someya.yoji@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan)

2012-09-15

Full text: JAEA has conducted the conceptual design study of blanket for a fusion DEMO reactor SlimCS. Considering DEMO specific requirements, we place emphasis on a blanket concept with durability to severe irradiation, ease of fabrication for mass production, operation temperature of blanket materials, and maintainability using remote handling equipment. This paper present a promising concept satisfying these requirements, which is characterized by minimized welding lines near the front, a simplified blanket interior consisting of cooling tubes and a mixed pebble bed of breeder and neutron multiplier, and approximately the same outlet temperature for all blanket modules. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in tritium breeding ratio even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production, which will facilitate the access of remote handling equipment for replacement of the blanket modules and improve the access of diagnostics. (author)

Imploding-liner reactor nucleonic studies: the LINUS blanket

International Nuclear Information System (INIS)

Dudziak, D.J.

1977-09-01

Scoping nucleonic studies have been performed for a small imploding-liner fusion reactor concept. Tritium breeding ratio and time-dependent energy deposition rates were the primary parameters of interest in the study. Alloys of Pb and LiPb were considered for the liquid liner (blanket), and tritium breeding was found to be more than adequate with blankets less than 1 m thick. However, neutron leakages into the solid cylinder block surrounding the liquid liner are generally quite high, so considerable effort was concentrated on minimizing these values. Time-dependent calculations reveal that 89% of the energy is deposited in the blanket within 2 μs. Thus, LINUS's blanket should remain intact for the requisite neutron and gamma-ray lifetimes

Muon catalyzed fusion - fission reactor driven by a recirculating beam

International Nuclear Information System (INIS)

Eliezer, S.; Tajima, T.; Rosenbluth, M.N.

1986-01-01

The recent experimentally inferred value of multiplicity of fusion of deuterium and tritium catalyzed by muons has rekindled interest in its application to reactors. Since the main energy expended is in pion (and consequent muon) productions, we try to minimize the pion loss by magnetically confining pions where they are created. Although it appears at this moment not possible to achieve energy gain by pure fusion, it is possible to gain energy by combining catalyzed fusion with fission blankets. We present two new ideas that improve the muon fusion reactor concept. The first idea is to combine the target, the converter of pions into muons, and the synthesizer into one (the synergetic concept). This is accomplished by injecting a tritium or deuterium beam of 1 GeV/nucleon into DT fuel contained in a magnetic mirror. The confined pions slow down and decay into muons, which are confined in the fuel causing little muon loss. The necessary quantity of tritium to keep the reactor viable has been derived. The second idea is that the beam passing through the target is collected for reuse and recirculated, while the strongly interacted portion of the beam is directed to electronuclear blankets. The present concepts are based on known technologies and on known physical processes and data. 29 refs., 6 figs., 4 tabs

Inverse kinetics for subcritical systems with external neutron source

International Nuclear Information System (INIS)

Carvalho Gonçalves, Wemerson de; Martinez, Aquilino Senra; Carvalho da Silva, Fernando

2017-01-01

Highlights: • It was developed formalism for reactivity calculation. • The importance function is related to the system subcriticality. • The importance function is also related with the value of the external source. • The equations were analyzed for seven different levels of sub criticality. • The results are physically consistent with others formalism discussed in the paper. - Abstract: Nuclear reactor reactivity is one of the most important properties since it is directly related to the reactor control during the power operation. This reactivity is influenced by the neutron behavior in the reactor core. The time-dependent neutrons behavior in response to any change in material composition is important for the reactor operation safety. Transient changes may occur during the reactor startup or shutdown and due to accidental disturbances of the reactor operation. Therefore, it is very important to predict the time-dependent neutron behavior population induced by changes in neutron multiplication. Reactivity determination in subcritical systems driven by an external neutron source can be obtained through the solution of the inverse kinetics equation for subcritical nuclear reactors. The main purpose of this paper is to find the solution of the inverse kinetics equation the main purpose of this paper is to device the inverse kinetics equations for subcritical systems based in a previous paper published by the authors (Gonçalves et al., 2015) and by (Gandini and Salvatores, 2002; Dulla et al., 2006). The solutions of those equations were also obtained. Formulations presented in this paper were tested for seven different values of k eff with external neutrons source constant in time and for a powers ratio varying exponentially over time.

Parametric systems analysis for tandem mirror hybrids

International Nuclear Information System (INIS)

Lee, J.D.; Chapin, D.L.; Chi, J.W.H.

1980-09-01

Fusion fission systems, consisting of fissile producing fusion hybrids combining a tandem mirror fusion driver with various blanket types and net fissile consuming LWR's, have been modeled and analyzed parametrically. Analysis to date indicates that hybrids can be competitive with mined uranium when U 3 O 8 cost is about 100 $/lb., adding less than 25% to present day cost of power from LWR's. Of the three blanket types considered, uranium fast fission (UFF), thorium fast fission (ThFF), and thorium fission supressed (ThFS), the ThFS blanket has a modest economic advantage under most conditions but has higher support ratios and potential safety advantages under all conditions

Fusion-reactor blanket-material safety-compatibility studies

International Nuclear Information System (INIS)

Jeppson, D.W.; Muhlestein, L.D.; Keough, R.F.; Cohen, S.

1982-11-01

Blanket material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Blanket material safety compatibility studies are being conducted to identify and characterize blanket-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate that : (1) ternary oxides (LiAlO 2 , Li 2 ZrO 3 , Li 2 SiO 3 , Li 4 SiO 4 and LiTiO 3 ) at postulated blanket operating temperatures are compatible with water coolant, while liquid lithium and Li 7 Pb 2 alloy reactions with water generate heat, aerosol and hydrogen; (2) lithium oxide and Li 17 Pb 83 alloy react mildly with water requiring special precautions to control hydrogen release; (3) liquid lithium reacts substantially, while Li 17 Pb 83 alloy reacts mildly with concrete to produce hydrogen; and (4) liquid lithium-air reactions present some major safety concerns

Source driven breeding thermal power reactors

International Nuclear Information System (INIS)

Greenspan, E.; Schneider, A.; Misulovin, A.; Gilai, D.; Levin, P.; Ben-Gurion Univ. of the Negev, Beersheba

1978-03-01

Improvements in the performance of fission power reactors made possible by designing them subcritical driven by D-T neutron sources are investigated. Light-water thermal systems are found to be most promising, neutronically and energetically, for the source driven mode of operation. The range of performance characteristics expected from breeding Light Water Hybrid Reactors (LWHR) is defined. Several promising types of LWHR blankets are identified. Options opened for the nuclear energy strategy by four types of the LWHRs are examined, and the potential contribution of these LWHRs to the nuclear energy economy are discussed. The power systems based on these LWHRs are found to enable a high utilization of the energy content of the uranium resources in all forms available - including depleted uranium and spent fuel from LWRs, while being free from the need for uranium enrichment and plutonium separation capabilities. (author)

ADS based on NaF-PbF2 molten salt

International Nuclear Information System (INIS)

Volk, V.I.; Vakhrushin, A.Yu.; Kwaratzkheli, A.Yu.; Konev, V.N.; Kochurov, B.P.; Shvedov, O.V.

1999-01-01

The neutron-physical parameters of an accelerator driven system (ADS) with a proton accelerator feeding a sub-critical molten salt blanket are investigated. The installation is designed for the production of electric power, involving thorium in a fuel cycle, transmutation of fission products and actinides. It is supposed to use fluoride salt composition 66PbF 2 -34NaF with addition of heavy elements (Th, Np, Pu and minor actinides) as the material of fuel, coolant and target. The thermal power of this ADS is 2000 MW. The current of the 1 GeV proton beam is 29 mA. The investigations are carried out for the following fuel cycles: the plutonium one, the burning of Np and minor actinides and the plutonium-thorium cycle. The balances of nuclides systems under supposition of its continuous operation during 20 years are presented [ru

European DEMO BOT solid breeder blanket

International Nuclear Information System (INIS)

Dalle Donne, M.

1994-11-01

The BOT (Breeder Outside Tube) Solid Breeder Blanket for a fusion DEMO reactor is presented. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. In the paper the reference blanket design and external loops are described as well as the results of the theoretical and experimental work in the fields of neutronics, thermohydraulics, mechanical stresses, tritium control and extraction, development and irradiation of the ceramic breeder material, beryllium development, ferromagnetic forces caused by disruptions, safety and reliability. An outlook is given on the remaining open questions and on the required R and D program. (orig.) [de

Compact D-D Neutron Source-Driven Subcritical Multiplier and Beam-Shaping Assembly for Boron Neutron Capture Therapy

International Nuclear Information System (INIS)

Ganda, Francesco; Vujic, Jasmina; Greenspan, Ehud; Leung, Ka-Ngo

2010-01-01

This work assesses the feasibility of using a small, safe, and inexpensive keff 0.98 subcritical fission assembly [subcritical neutron multiplier (SCM)] to amplify the treatment neutron beam intensity attainable from a compact deuterium-deuterium (D-D) fusion neutron source delivering [approximately]1012 n/s. The objective is to reduce the treatment time for deep-seated brain tumors to [approximately]1 h. The paper describes the optimal SCM design and two optimal beam-shaping assemblies (BSAs) - one designed to maximize the dose rate and the other designed to maximize the total dose that can be delivered to a deep-seated tumor. The neutron beam intensity amplification achieved with the optimized SCM and BSA results in an increase in the treatment dose rate by a factor of 18: from 0.56 Gy/h without the SCM to 10.1 Gy/h. The entire SCM is encased in an aluminum structure. The total amount of 20% enriched uranium required for the SCM is 8.5 kg, and the cost (not including fabrication) is estimated to be less than $60,000. The SCM power level is estimated at 400 W when driven by a 1012 n/s D-D neutron source. This translates into consumption of only [approximately]0.6% of the initially loaded 235U atoms during 50 years of continuous operation and implies that the SCM could operate continuously for the entire lifetime of the facility without refueling. Cooling the SCM does not pose a challenge; it may be accomplished by natural circulation as the maximum heat flux is only 0.034 W/cm2.

A microfluidic sub-critical water extraction instrument

Science.gov (United States)

Sherrit, Stewart; Noell, Aaron C.; Fisher, Anita; Lee, Mike C.; Takano, Nobuyuki; Bao, Xiaoqi; Kutzer, Thomas C.; Grunthaner, Frank

2017-11-01

This article discusses a microfluidic subcritical water extraction (SCWE) chip for autonomous extraction of amino acids from astrobiologically interesting samples. The microfluidic instrument is composed of three major components. These include a mixing chamber where the soil sample is mixed and agitated with the solvent (water), a subcritical water extraction chamber where the sample is sealed with a freeze valve at the chip inlet after a vapor bubble is injected into the inlet channels to ensure the pressure in the chip is in equilibrium with the vapor pressure and the slurry is then heated to ≤200 °C in the SCWE chamber, and a filter or settling chamber where the slurry is pumped to after extraction. The extraction yield of the microfluidic SCWE chip process ranged from 50% compared to acid hydrolysis and 80%-100% compared to a benchtop microwave SCWE for low biomass samples.

Beryllium R and D for blanket application

Energy Technology Data Exchange (ETDEWEB)

Dalle Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Longhurst, G.R. [Idaho National Engineering Lab., Idaho Falls (United States); Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

1998-10-01

The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.) 29 refs.

Beryllium R and D for blanket application

International Nuclear Information System (INIS)

Dalle Donne, M.; Scaffidi-Argentina, F.; Kawamura, H.

1998-01-01

The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.)

Blanket safety by GEMSAFE methodology

International Nuclear Information System (INIS)

Sawada, Tetsuo; Saito, Masaki

2001-01-01

General Methodology of Safety Analysis and Evaluation for Fusion Energy Systems (GEMSAFE) has been applied to a number of fusion system designs, such as R-tokamak, Fusion Experimental Reactor (FER), and the International Thermonuclear Experimental Reactor (ITER) designs in the both stages of Conceptual Design Activities (CDA) and Engineering Design Activities (EDA). Though the major objective of GEMSAFE is to reasonably select design basis events (DBEs) it is also useful to elucidate related safety functions as well as requirements to ensure its safety. In this paper, we apply the methodology to fusion systems with future tritium breeding blankets and make clear which points of the system should be of concern from safety ensuring point of view. In this context, we have obtained five DBEs that are related to the blanket system. We have also clarified the safety functions required to prevent accident propagations initiated by those blanket-specific DBEs. The outline of the methodology is also reviewed. (author)

Production of rare sugars from common sugars in subcritical aqueous ethanol.

Science.gov (United States)

Gao, Da-Ming; Kobayashi, Takashi; Adachi, Shuji

2015-05-15

A new isomerization reaction was developed to synthesize rare ketoses. D-tagatose, D-xylulose, and D-ribulose were obtained in the maximum yields of 24%, 38%, and 40%, respectively, from the corresponding aldoses, D-galactose, D-xylose, and D-ribose, by treating the aldoses with 80% (v/v) subcritical aqueous ethanol at 180°C. The maximum productivity of D-tagatose was ca. 80 g/(Lh). Increasing the concentration of ethanol significantly increased the isomerization of D-galactose. Variation in the reaction temperature did not significantly affect the production of D-tagatose from D-galactose. Subcritical aqueous ethanol converted both 2,3-threo and 2,3-erythro aldoses to the corresponding C-2 ketoses in high yields. Thus, the treatment of common aldoses in subcritical aqueous ethanol can be regarded as a new method to synthesize the corresponding rare sugars. Copyright © 2014 Elsevier Ltd. All rights reserved.

INTOR first wall/blanket/shield activity

International Nuclear Information System (INIS)

Gohar, Y.; Billone, M.C.; Cha, Y.S.; Finn, P.A.; Hassanein, A.M.; Liu, Y.Y.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.

1986-01-01

The main emphasis of the INTOR first wall/blanket/shield (FWBS) during this period has been upon the tritium breeding issues. The objective is to develop a FWBS concept which produces the tritium requirement for INTOR operation and uses a small fraction of the first wall surface area. The FWBS is constrained by the dimensions of the reference design and the protection criteria required for different reactor components. The blanket extrapolation to commercial power reactor conditions and the proper temperature for power extraction have been sacrificed to achieve the highest possible local tritium breeding ratio (TBR). In addition, several other factors that have been considered in the blanket survey study include safety, reliability, lifetime fluence, number of burn cycles, simplicity, cost, and development issues. The implications of different tritium supply scenarios were discussed from the cost and availability for INTOR conditions. A wide variety of blanket options was explored in a preliminary way to determine feasibility and to see if they can satisfy the INTOR conditions. This survey and related issues are summarized in this report. Also discussed are material design requirements, thermal hydraulic considerations, structure analyses, tritium permeation through the first wall into the coolant, and tritium inventory

ITER driver blanket, European Community design

International Nuclear Information System (INIS)

Simbolotti, G.; Zampaglione, V.; Ferrari, M.; Gallina, M.; Mazzone, G.; Nardi, C.; Petrizzi, L.; Rado, V.; Violante, V.; Daenner, W.; Lorenzetto, P.; Gierszewski, P.; Grattarola, M.; Rosatelli, F.; Secolo, F.; Zacchia, F.; Caira, M.; Sorabella, L.

1993-01-01

Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly afftect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. With the framework of the ITER Conceptual Design Activities (CDA, 1988-1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper. (orig.)

Feasibility study of a fission-suppressed tokamak fusion breeder

International Nuclear Information System (INIS)

Moir, R.W.; Lee, J.D.; Neef, W.S.

1984-12-01

The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m 2 and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of 233 U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U 3 O 8 depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management

Non-destructive assay of EBR-II blanket elements using resonance transmission analysis

International Nuclear Information System (INIS)

Klann, R.T.; Poenitz, W.P.

1998-01-01

Resonance transmission analysis utilizing a faltered reactor beam was examined as a means of determining the 239 Pu content in Experimental Breeder Reactor-II depleted uranium blanket elements. The technique uses cadmium and gadolinium falters along with a 239 Pu fission chamber to isolate the 0.3 eV resonance in 239 Pu. In the energy range of this resonance (0.1 eV to 0.5 ev), the total microscopic cross-section of 239 Pu is significantly greater than the cross-sections of 238 U and 235 U. This large difference allows small changes in the 239 Pu content of a sample to result in large changes in the mass signal response. Tests with small stacks of depleted uranium and 239 Pu foils indicate a significant change in response based on the 239 Pu content of the foil stack. In addition, the tests indicate good agreement between the measured and predicted values of 239 Pu up to approximately two weight percent

Measurement of Fission Product Yields from Fast-Neutron Fission

Science.gov (United States)

Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Henderson, R.; Kenneally, J.; Macri, R.; McNabb, D.; Ryan, C.; Sheets, S.; Stoyer, M. A.; Tonchev, A. P.; Bhatia, C.; Bhike, M.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.

2014-09-01

One of the aims of the Stockpile Stewardship Program is a reduction of the uncertainties on fission data used for analyzing nuclear test data [1,2]. Fission products such as 147Nd are convenient for determining fission yields because of their relatively high yield per fission (about 2%) and long half-life (10.98 days). A scientific program for measuring fission product yields from 235U,238U and 239Pu targets as a function of bombarding neutron energy (0.1 to 15 MeV) is currently underway using monoenergetic neutron beams produced at the 10 MV Tandem Accelerator at TUNL. Dual-fission chambers are used to determine the rate of fission in targets during activation. Activated targets are counted in highly shielded HPGe detectors over a period of several weeks to identify decaying fission products. To date, data have been collected at neutron bombarding energies 4.6, 9.0, 14.5 and 14.8 MeV. Experimental methods and data reduction techniques are discussed, and some preliminary results are presented.

Fusion blanket testing in MFTF-α + T

International Nuclear Information System (INIS)

Kleefeldt, K.

1985-01-01

The Mirror Fusion Test Facility-α + T (MFTF-α + T) is an upgraded version of the current MFTF-B test facility at Lawrence Livermore National Laboratory, and is designed for near-term fusion-technology-integrated tests at a neutron flux of 2 MW/m 2 . Currently, the fusion community is screening blanket and related issues to determine which ones can be addressed using MFTF-α + T. In this work, the minimum testing needs to address these issues are identified for the liquid-metal-cooled blanket and the solid-breeder blanket. Based on the testing needs and on the MFTF-α + T capability, a test plan is proposed for three options; each option covers a six to seven year testing phase. The options reflect the unresolved question of whether to place the research and development (R and D) emphasis on liquid-metal or solid-breeder blankets. In each case, most of the issues discussed can be addressed to a reasonable extent in MFTF-α+T

Pulsed activation analyses of the ITER blanket design options considered in the blanket trade-off study

International Nuclear Information System (INIS)

Wang, Q.; Henderson, D.L.

1995-01-01

Pulsed activation calculations have been performed on two blanket options considered as part of the ITER home team blanket trade-off study. The objective was to compare the activity, afterheat and waste disposal rating (WDR) results of a composite blanket-shield design for the continuous operation approximation to a pulsed operation case to determine whether the differences are at most the duty factor as predicted by the two nuclide chain model. Up to a cooling period of 100 years, the pulsed activity and afterheat values were below the continuous oepration results and well within (except for one afterheat value) the maximum deviation predicted by the two nuclide chain model. No differences in the WDR values were noted as they are, to a large extent, based on long-lived nuclides which are insensitive to short-term changes in the operation history. (orig.)

Measurement of kinetic parameters in the fast subcritical core MASURCA

International Nuclear Information System (INIS)

Baeten, Peter; Abderrahim, Hamid Aiet

2004-01-01

In the MUSE shared cost action of the European Fifth Framework Program measurements have been performed to investigate the neutronic behavior of the fast subcritical core MASURCA coupled with the GENEPI accelerator. The aim is to examine the applicability of different measurement techniques for the determination of the main kinetic parameters. The measurement of Rossi-alpha distributions, recorded with the accelerator turned off, showed that the analysis of the obtained distributions is feasible for deep subcritical levels, but with strongly deteriorated statistics. From Rossi-alpha distributions, recorded with the pulsed neutron source in operation, the alpha decay constant was easily derived due to good statistics on the correlated signal resulting from the strong intensity of the neutron pulse. When applying the pulsed neutron source analysis, the reactivity (in dollars) together with the ratio of the mean neutron lifetime l and the effective delayed neutron fraction β eff is immediately derived. Although these first results are very promising, further measurements are needed to qualify the method at larger subcritical levels which are representative for future ADS

Axial blanket enrichment optimization of the NPP Krsko fuel

International Nuclear Information System (INIS)

Kromar, M.; Kurincic, B.

2001-01-01

In this paper optimal axial blanket enrichment of the NPP Krsko fuel is investigated. Since the optimization is dictated by economic categories that can significantly vary in time, two step approach is applied. In the first step simple relationship between the equivalent change in enrichment of axial blankets and central fuel region is established. The relationship is afterwards processed with economic criteria and constraints to obtain optimal axial blanket enrichment. In the analysis realistic NPP Krsko conditions are considered. Except for the fuel enrichment all other fuel characteristics are the same as in the fuel used in the few most recent cycles. A typical reload cycle after the plant power uprate is examined. Analysis has shown that the current blanket enrichment is close to the optimal. Blanket enrichment reduction results in an approximately 100 000 US$ savings per fuel cycle.(author)

K/sub infinity/-meter concept verified via subcritical-critical TRIGA experiments

International Nuclear Information System (INIS)

Ocampo Mansilla, H.

1983-01-01

This work presents a technique for building a device to measure the k/sub infinity/ of a spent nuclear fuel assembly discharged from the core of a nuclear power plant. The device, called a k/sub infinity/-meter, consists of a cross-shaped subcritical assembly, two artificial neutron sources, and two separate neutron counting systems. The central position of the subcritical assembly is used to measure k/sub infinity/ of the spent fuel assembly. The initial subcritical assembly is calibrated to determine its k/sub eff/ and verify the assigned k/sub infinity/ of a selected fuel assembly placed in the central position. Count rates are taken with the fuel assembly of known k/sub infinity/'s placed in the central position and then repeated with a fuel assembly of unknown k/sub infinity/ placed in the central position. The count rate ratio of the unknown fuel assembly to the known fuel assembly is used to determine the k/sub infinity/ of the unknown fuel assembly. The k/sub infinity/ of the unknown fuel assembly is represented as a polynomial function of the count rate ratios. The coefficients of the polynomial equation are determined using the neutronic codes LEOPARD and EXTERMINATOR-II. The analytical approach has been validated by performing several subcritical/critical experiments, using the Penn State Breazeale TRIGA Reactor (PSBR), and comparing the experimental results with the calculations

Trade-off study of liquid-metal self-cooled blankets

International Nuclear Information System (INIS)

Gohar, Y.

1986-01-01

A trade-off study of liquid-metal self-cooled blankets was carried out to define the performance of these blankets with respect to the main functions in a fusion reactor, and to determine the potential to operate at the maximum possible values of the performance parameters. The main purpose is to improve the reactor economics by maximizing the blanket energy multiplication factor, reduce the capital cost of the reactor, and satisfy the design requirements. The main parameters during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the 6 Li enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, the impact of different reactor design choices on the performance parameters was analyzed. The effect of the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, the coolant choice for the nonbreeding inboard blanket, and the neutron source distribution were part of the trade-off study. In addition, tritium breeding benchmark calculations were performed to study the impact of the use of different transport codes and nuclear data libraries. The importance and the negative effect of high TBR on the energy multiplication motivated the benchmark calculations

Design and analysis of ITER shield blanket

Energy Technology Data Exchange (ETDEWEB)

Ohmori, Junji; Hatano, Toshihisa; Ezato, Kouichiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1998-12-01

This report includes electromagnetic analyses for ITER shielding blanket modules, fabrication methods for the blanket modules and the back plate, the design and the fabrication methods for port limiter have been investigated. Studies on the runaway electron impact for Be armor have been also performed. (J.P.N.)

Non-electrical uses of thermal energy generated in the production of fissile fuel in fusion--fission reactors: a comparative economic parametric analysis for a hybrid with or without synthetic fuel production

International Nuclear Information System (INIS)

Tai, A.S.; Krakowski, R.A.

1979-01-01

A parametric analysis has been carried out for testing the sensitivity of the synfuel production cost in relation to crucial economic and technologic quantities (investment costs of hybrid and synfuel plant, energy multiplication of the fission blanket, recirculating power fraction of the fusion driver, etc.). In addition, a minimum synfuel selling price has been evaluated, from which the fission--fusion--synfuel complex brings about a higher economic benefit than does the fusion--fission hybrid entirely devoted to fissile-fuel and electricity generation. Assuming an electricity cost of 2.7 cents/kWh, an annual investment cost per power unit of 4.2 to 6 $/GJ (132 to 189 k$/MWty) for the fission--fusion complex and 1.5 to 3 $/GJ (47 to 95 k$/MWty) for the synfuel plant, the synfuel production net cost (i.e., revenue = cost) varies between 6.5 and 8.6 $/GJ. These costs can compete with those obtained by other processes (natural gas reforming, resid partial oxidation, coal gasification, nuclear fission, solar electrolysis, etc.). This study points out a potential use of the fusion--fission hybrid other than fissile-fuel and electricity generation

Processing and waste disposal needs for fusion breeder blankets system

International Nuclear Information System (INIS)

Finn, P.A.; Vogler, S.

1988-01-01

We evaluated the waste disposal and recycling requirements for two types of fusion breeder blanket (solid and liquid). The goal was to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under U.S. Nuclear Regulatory Commission regulations. Described in this paper are the radionuclides expected in fusion blanket materials, plans for reprocessing and disposal of blanket components, and estimates for the operating costs involved in waste disposal. (orig.)

[Fission product yields of 60 fissioning reactions]. Final report

International Nuclear Information System (INIS)

Rider, B.F.

1995-01-01

In keeping with the statement of work, I have examined the fission product yields of 60 fissioning reactions. In co-authorship with the UTR (University Technical Representative) Talmadge R. England ''Evaluation and Compilation of Fission Product Yields 1993,'' LA-UR-94-3106(ENDF-349) October, (1994) was published. This is an evaluated set of fission product Yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

Epoxy blanket protects milled part during explosive forming

Science.gov (United States)

1966-01-01

Epoxy blanket protects chemically milled or machined sections of large, complex structural parts during explosive forming. The blanket uniformly covers all exposed surfaces and fills any voids to support and protect the entire part.

Impact of fusion-fission hybrids on world nuclear future

International Nuclear Information System (INIS)

Abdel-Khalick, S.; Jansen, P.; Kessler, G.; Klumpp, P.

1980-08-01

An investigation has been conducted to examine the impact of fusion-fission hybrids on world nuclear future. The primary objectives of this investigation have been: (1) to determine whether hybrids can allow us to meet the projected nuclear component of the world energy demand within current estimates of uranium resources without fast breeders, and (2) to identify the preferred hybrid concept from a resource standpoint. The results indicate that hybrids have the potential to lower the world uranium demand to values well below the resource base. However, the time window for hybrid introduction is quite near and narrow (2000-2020). If historical market penetration rates are assumed, the demand will not be met within the resource base unless hybrids are coupled to the breeders. The results also indicate that from a resource standpoint hybrids which breed their own tritium and have a low blanket energy multiplication are preferable. (orig.) [de

Contained fission explosion breeder reactor system

International Nuclear Information System (INIS)

Juhl, N.H.; Marwick, E.F.

1983-01-01

A reactor system for producing useful thermal energy and valuable isotopes, such as plutonium-239, uranium-233, and/or tritium, in which a pair of sub-critical masses of fissile and fertile actinide slugs are propelled into an ellipsoidal pressure vessel. The propelled slugs intercept near the center of the chamber where the concurring slugs become a more than prompt configuration thereby producing a fission explosion. Re-useable accelerating mechanisms are provided external of the vessel for propelling the slugs at predetermined time intervals into the vessel. A working fluid of lean molten metal slurry is injected into the chamber prior to each explosion for the attenuation of the explosion's effects, for the protection of the chamber's walls, and for the absorbtion of thermal energy and debris from the explosion. The working fluid is injected into the chamber in a pattern so as not to interfere with the flight paths of the slugs and to maximize the concentration of working fluid near the chamber's center. The heated working fluid is drained from the vessel and is used to perform useful work. Most of the debris from the explosion is collected as precipitate and is used for the manufacture of new slugs

Subcriticality calculation in nuclear reactors with external neutron sources

Energy Technology Data Exchange (ETDEWEB)

Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mails: asilva@con.ufrj.br; aquilino@lmp.ufrj.br; fernando@con.ufrj.br

2007-07-01

The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

Subcriticality calculation in nuclear reactors with external neutron sources

International Nuclear Information System (INIS)

Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

2007-01-01

The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

18 CFR 284.303 - OCS blanket certificates.

Science.gov (United States)

2010-04-01

... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false OCS blanket certificates. 284.303 Section 284.303 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... Pipelines on Behalf of Others § 284.303 OCS blanket certificates. Every OCS pipeline [as that term is...

Promotion or suppression of glucose isomerization in subcritical aqueous straight- and branched-chain alcohols.

Science.gov (United States)

Gao, Da-Ming; Kobayashi, Takashi; Adachi, Shuji

2015-01-01

The influence of water-miscible alcohols (methanol, 1-propanol, 2-propanol, and t-butyl alcohol) on the isomerization of glucose to fructose and mannose was investigated under subcritical aqueous conditions (180-200 °C). Primary and secondary alcohols promoted the conversion and isomerization of glucose to afford fructose and mannose with high and low selectivity, respectively. On the other hand, the decomposition (side-reaction) of glucose was suppressed in the presence of the primary and secondary alcohols compared with that in subcritical water. The yield of fructose increased with increasing concentration of the primary and secondary alcohols, and the species of the primary and secondary alcohols tested had little effect on the isomerization behavior of glucose. In contrast, the isomerization of glucose was suppressed in subcritical aqueous t-butyl alcohol. Both the conversion of glucose and the yield of fructose decreased with increasing concentration of t-butyl alcohol. In addition, mannose was not detected in reactions using subcritical aqueous t-butyl alcohol.

Noise method for monitoring the sub-criticality in accelerator driven systems

International Nuclear Information System (INIS)

Rugama, Y.; Munoz-Cobo, J.L.; Valentine, T.E.; Mihalczo, J.T.; Perez, R.B.; Perez-Navarro, A.

2001-01-01

In this paper, an absolute measurements technique for the sub-criticality determination is presented. The development of ADS, requires of methods to monitor and control the sub-criticality of this kind of systems, without interfering it's normal operation mode. This method is based on the Stochastic Neutron and Photon Transport Theory developed by Munoz-Cobo et al., and which can be implemented in presently available neutron transport codes. As a by-product of the methodology a monitoring measurement technique has been developed and verified using two coupled Monte Carlo programs. The spallation collisions and the high-energy transport are simulated with LAHET. The neutrons transports with energies less than 20 MeV and the estimation of the count statistics for neutron and/or gamma ray counters in fissile systems, is simulated with MCNP-DSP. It is possible to get the kinetics parameters and the k eff value of the sub-critical system through the analysis of the counter detectors. (author)

Japanese contributions to the Japan-US workshop on blanket design/technology

International Nuclear Information System (INIS)

Tone, Tatsuzo; Seki, Yasushi; Minato, Akio; Kobayashi, Takeshi; Mori, Seiji; Kawasaki, Hiromitsu; Sumita, Kenji.

1983-02-01

This report describes Japanese papers presented at the Japan-US Workshop on Blanket Design/Technology which was held at Argonne National Laboratory, November 10 - 11, 1982. Overview of Fusion Experimental Reactor (FER), JAERI's activities related to first wall/blanket/shield, summary of FER blanket and its technology development issues and summary of activities at universities on fusion reactor blanket engineering are covered. (author)

Cassette blanket and vacuum building: key elements in fusion reactor maintenance

International Nuclear Information System (INIS)

Werner, R.W.

1977-01-01

The integration of two concepts important to fusion power reactors is discussed. The first concept is the vacuum building which improves upon the current fusion reactor designs. The second concept, the use of the cassette blanket within the vacuum building environment, introduces four major improvements in blanket design: cassette blanket module, zoning concept, rectangular blanket concept, and internal tritium recovery

Joint Markov Blankets in Feature Sets Extracted from Wavelet Packet Decompositions

Directory of Open Access Journals (Sweden)

Gert Van Dijck

2011-07-01

Full Text Available Since two decades, wavelet packet decompositions have been shown effective as a generic approach to feature extraction from time series and images for the prediction of a target variable. Redundancies exist between the wavelet coefficients and between the energy features that are derived from the wavelet coefficients. We assess these redundancies in wavelet packet decompositions by means of the Markov blanket filtering theory. We introduce the concept of joint Markov blankets. It is shown that joint Markov blankets are a natural extension of Markov blankets, which are defined for single features, to a set of features. We show that these joint Markov blankets exist in feature sets consisting of the wavelet coefficients. Furthermore, we prove that wavelet energy features from the highest frequency resolution level form a joint Markov blanket for all other wavelet energy features. The joint Markov blanket theory indicates that one can expect an increase of classification accuracy with the increase of the frequency resolution level of the energy features.

Neutronics analysis for aqueous self-cooled fusion reactor blankets

International Nuclear Information System (INIS)

Varsamis, G.; Embrechts, M.J.; Jaffa, R.; Steiner, D.; Deutsch, L.; Gierszewski, P.

1986-06-01

The tritium breeding performance of several Aqueous Self-Cooled Blanket (ASCB) configurations for fusion reactors has been evaluated. The ASCB concept employs small amounts of lithium compound dissolved in light or heavy water to serve as both coolant and breeding medium. The inherent simplicity of this concept allows the development of blankets with minimal technological risk. The tritium breeding performance of the ASCB concept is a critical issue for this family of blankets. Contrary to conventional blanket designs there will be a significant contribution to the tritium breeding ratio (TBR) in the water coolant/breeder of duct shields, and the 3-D TBR will therefore be similar to the 1-D TBR. The tritium breeding performance of an ASCB for a MARS-like-tandem reactor and an ASCB based breeding-shield for the Next European Torus (NET) are assessed. Two design options for the MARS-like blanket are discussed. One design employs a vanadium first wall, and zircaloy for the structural material. The trade-offs between light water and heavy water cooling options for this zircaloy blanket are discussed. The second design option for MARS relies on the use of a vanadium alloy as the stuctural material, and heavy water as the coolant. It is demonstrated that both design options lead to low-activation blankets that allow class C burial. The breeder-shield for NET consists of a water-cooled stainless steel shield

EU DEMO blanket concepts safety assessment. Final report of Working Group 6a of the Blanket Concept Selection Exercise

International Nuclear Information System (INIS)

Kleefeldt, K.; Porfiri, T.

1996-06-01

The European Union has been engaged since 1989 in a programme to develop tritium breeding blankets for application in a fusion power reactor. There are four blanket concepts under development. Two of them use lithium ceramics, the other two concepts employ an eutectic lead-lithium alloy (Pb-17Li) as breeder material. The two most promising concepts were to select in 1995 for further development. In order to prepare the selection, a Blanket Concept Selection Exercise (BCSE) has been inititated by the participating associations under the auspices of the European Commission. This BCSE has been performed in 14 working groups which, in a comparative evaluation of the four blanket concepts, addressed specific fields. The working group safety addressed the safety implications. This report describes the methodology adopted, the safety issues identified, their comparative evaluation for the four concepts, and the results and conclusions of the working group to be entered into the overall evaluation. There, the results from all 14 working groups have been combined to yield a final ranking as a basis for the selection. In summary, the safety assessment showed that the four European blanket concepts can be considered as equivalent in terms of the safety rating adopted, each concept, however, rendering safety concerns of different quality in different areas which are substantiated in this report. (orig.) [de

Tritium transport in the water cooled Pb-17Li blanket concept of DEMO

International Nuclear Information System (INIS)

Reiter, F.; Tominetti, S.; Perujo, A.

1992-01-01

The code TIRP has been used to calculate the time dependence of tritium inventory and tritium permeation into the coolant and into the first wall boxes in the water cooled Pb-17Li blanket concept of DEMO. The calculations have been performed for the martensitic steel MANET and the austenitic steel AISI 316L as blanket structure materials, for water or helium cooling and for convective or no motion of the liquid breeder in the blanket. Tritium inventories are rather low in blankets with MANET structure and higher in those with AISI 316L structure. Tritium permeation rates are too high in both blankets. Further calculations on tritium inventory and permeation are therefore presented for blankets with TiC permeation barriers of 1 μm thickness on various surfaces of the blanket structure and for blankets with any permeation barriers in function of their thickness, tritium diffusivities, tritium surface recombination rates and atomic densities. These last calculations have been performed for a blanket with coatings on the outer surfaces of the blanket and with a tritium residence time of 10 4 s and for a blanket with coatings on both sides of the cooling tubes and stagnant Pb-17Li in the blanket. The second case for a blanket with MANET structure presents a very interesting solution for tritium recovery by permeation into and pumping from the first wall boxes. (orig.)

Fusion breeder sphere - PAC blanket design

International Nuclear Information System (INIS)

Sullivan, J.D.; Palmer, B.J.F.

1987-11-01

There is a considerable world-wide effort directed toward the production of materials for fusion reactors. Many ceramic fabrication groups are working on making lithium ceramics in a variety of forms, to be incorporated into the tritium breeding blanket which will surround the fusion reactor. Current blanket designs include ceramic in either monolithic or packed sphere bed (sphere-pac) forms. The major thrust at AECL is the production of lithium aluminate spheres to be incorporated in a sphere-pac bed. Contemporary studies on breeder blanket design offer little insight into the requirements on the sizes of the spheres. This study examined the parameters which determine the properties of pressure drop and coolant requirements. It was determined that an optimised sphere-pac bed would be composed of two diameters of spheres: 75 weight % at 3 mm and 25 weight % at 0.3 mm

Effect of reactor size on the breeding economics of LMFBR blankets

International Nuclear Information System (INIS)

Tagishi, A.; Driscoll, M.J.

1975-02-01

The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MW(e) were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's

A new formulation for the importance function in the kinetics of subcritical reactors

International Nuclear Information System (INIS)

Silva, Cristiano da; Senra Martinez, Aquilino; Carvalho da Silva, Fernando

2012-01-01

Highlights: ► In this paper we propose a new formulation for the importance function in the kinetics of subcritical systems. ► We analyze the relevance of an external neutron source for the subcritical interval 0.95 eff eff is the multiplication factor according to the physical properties of the nuclear reactor. For the purposes of validation of the proposed method we will use, as a reference method, the expansion in modes of the time-dependent neutron flux for the solution of the onedimensional diffusion equation. It will be presented results that demonstrate the precision of the proposed method when compared to the conventional point kinetic equations. The results show that the new point kinetic equations are rather precise in the subcriticality range considered.

Assessment of alkali metal coolants for the ITER blanket

International Nuclear Information System (INIS)

Natesan, K.; Reed, C.B.; Mattas, R.F.

1994-01-01

The blanket system is one of the most important components of a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The Blanket Comparison and Selection Study, conducted earlier, described the overall comparative performance of different blanket concepts, including liquid metal, molten salt, water, and helium. This paper will discuss the ITER requirements for a self-cooled blanket concept with liquid lithium and for indirectly cooled concepts that use other alkali metals such as NaK. The paper will address the thermodynamics of interactions between the liquid metals (i.e., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data will be used to assess the long-term performance of the first wall in a liquid metal environment

Analysis of ER string test thermally instrumented interconnect 80-K MLI blanket

International Nuclear Information System (INIS)

Daly, E.; Pletzer, R.

1992-04-01

An 80-K Multi Layer Insulation (MLI) blanket in the interconnect region between magnets DD0019 and DD0027 in the Fermi National Accelerator Laboratory (FNAL) ER string was instrumented with temperature sensors to obtain the steady-state temperature gradient through the blanket after string cooldown. A thermal model of the 80-K blanket assembly was constructed to analyze the steady-state temperature gradient data. Estimates of the heat flux through the 80-K MLI blanket assembly and predicted temperature gradients were calculated. The thermal behavior of the heavy polyethylene terapthalate (PET) cover layers separating the shield and inner blanket and inner and outer blankets was derived empirically from the data. The results of the analysis predict a heat flux of 0.363--0.453 W/m 2 based on the 11 sets of data. These flux values are 33--46% below the 80-K MLI blanket heat leak budget of 0.676 W/m 2 . The effective thermal resistance of the two heavy PET cover layers between the shield and inner blanket was found to be 2.1 times that of a single PET spacer layer, and the effective resistance of the two heavy PET cover layers between the inner blanket and outer blanket was found to be 7 times that of a single PET spacer layer. Based on these results, the 80-K MLI blanket assembly appears to be performing more than adequately to meet the 80-K static IR heat leak budget. However, these results should not be construed as a verification of the 80-K static IR heat leak, since no actual heat leak was measured. The results have been used to improve the empirically based model data in the 80-K MLI blanket thermal model, which has previously not included the effects of heavy PET cover layers on 80-K MLI blanket thermal performance

A review of fusion breeder blanket technology, part 1

International Nuclear Information System (INIS)

Jackson, D.P.; Selander, W.N.; Townes, B.M.

1985-01-01

This report presents the results of a study of fusion breeder blanket technology. It reviews the role of the breeder blanket, the current understanding of the scientific and engineering bases of liquid metal and solid breeder blankets and the programs now underway internationally to resolve the uncertainities in current knowledge. In view of existing national expertise and experience, a solid breeder R and D program for Canada is recommended

Criticality Analysis of SAMOP Subcritical Assembly

International Nuclear Information System (INIS)

Tegas-Sutondo; Syarip; Triwulan-Tjiptono

2005-01-01

A critically analysis has been performed for homogenous system of uranyl nitrate solution, as part of a preliminary design assessment on neutronic aspect of SAMOP sub-critical assembly. The analysis is intended to determine some critical parameters such as the minimum of critical dimension and critical mass for the desired concentration. As the basis of this analysis, it has been defined a fuel system with an enrichment of 20% for cylindrical geometry of both bare and graphite reflected of 30 cm thickness. The MCNP code has been utilized for this purpose, for variation of concentrations ranging from 150 g/l to 500 g/l. It is found that the best concentration giving the minimum geometrical dimension is around 400 g/l, for both the bare and reflected systems. Whilst the best one, of minimum critical mass is corresponding to the concentration of around 200 g/l with critical mass around 14.1 kg and 4.2 kg for the bare and reflected systems respectively. Based on the result of calculations, it is concluded that by taking into consideration of the critical limit, the SAMOP subcritical assembly is neutronically can be made. (author)

Ternary fission

International Nuclear Information System (INIS)

Wagemans, C.

1991-01-01

Since its discovery in 1946, light (charged) particle accompanied fission (ternary fission) has been extensively studied, for spontaneous as well as for induced fission reactions. The reason for this interest was twofold: the ternary particles being emitted in space and time close to the scission point were expected to supply information on the scission point configuration and the ternary fission process was an important source of helium, tritium, and hydrogen production in nuclear reactors, for which data were requested by the nuclear industry. Significant experimental progress has been realized with the advent of high-resolution detectors, powerful multiparameter data acquisition systems, and intense neutron and photon beams. As far as theory is concerned, the trajectory calculations (in which scission point parameters are deduced from the experimental observations) have been very much improved. An attempt was made to explain ternary particle emission in terms of a Plateau-Rayleigh hydrodynamical instability of a relatively long cylindrical neck or cylindrical nucleus. New results have also been obtained on the so-called open-quotes trueclose quotes ternary fission (fission in three about-equal fragments). The spontaneous emission of charged particles has also clearly been demonstrated in recent years. This chapter discusses the main characteristics of ternary fission, theoretical models, light particle emission probabilities, the dependence of the emission probabilities on experimental variables, light particle energy distributions, light particle angular distributions, correlations between light particle accompanied fission observables, open-quotes trueclose quotes ternary fission, and spontaneous emission of heavy ions. 143 refs., 18 figs., 8 tabs

FW/Blanket and vacuum vessel for RTO/RC ITER

International Nuclear Information System (INIS)

Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Iida, H.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Yamada, M.

2000-01-01

The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, ∼50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste

FW/Blanket and vacuum vessel for RTO/RC ITER

Energy Technology Data Exchange (ETDEWEB)

Ioki, K. E-mail: iokik@itereu.de; Barabash, V.; Cardella, A.; Elio, F.; Iida, H.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Yamada, M

2000-11-01

The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, {approx}50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste.

Fission level densities

International Nuclear Information System (INIS)

Maslov, V.M.

1998-01-01

Fission level densities (or fissioning nucleus level densities at fission saddle deformations) are required for statistical model calculations of actinide fission cross sections. Back-shifted Fermi-Gas Model, Constant Temperature Model and Generalized Superfluid Model (GSM) are widely used for the description of level densities at stable deformations. These models provide approximately identical level density description at excitations close to the neutron binding energy. It is at low excitation energies that they are discrepant, while this energy region is crucial for fission cross section calculations. A drawback of back-shifted Fermi gas model and traditional constant temperature model approaches is that it is difficult to include in a consistent way pair correlations, collective effects and shell effects. Pair, shell and collective properties of nucleus do not reduce just to the renormalization of level density parameter a, but influence the energy dependence of level densities. These effects turn out to be important because they seem to depend upon deformation of either equilibrium or saddle-point. These effects are easily introduced within GSM approach. Fission barriers are another key ingredients involved in the fission cross section calculations. Fission level density and barrier parameters are strongly interdependent. This is the reason for including fission barrier parameters along with the fission level densities in the Starter File. The recommended file is maslov.dat - fission barrier parameters. Recent version of actinide fission barrier data obtained in Obninsk (obninsk.dat) should only be considered as a guide for selection of initial parameters. These data are included in the Starter File, together with the fission barrier parameters recommended by CNDC (beijing.dat), for completeness. (author)

Progress in blanket designs using SiCf/SiC composites

International Nuclear Information System (INIS)

Giancarli, L.; Golfier, H.; Nishio, S.; Raffray, R.; Wong, C.; Yamada, R.

2002-01-01

This paper summarizes the most recent design activities concerning the use of SiC f /SiC composite as structural material for fusion power reactor breeding blanket. Several studies have been performed in the past. The most recent proposals are the TAURO blanket concept in the European Union, the ARIES-AT concept in the US, and DREAM concept in Japan. The first two concepts are self-cooled lithium-lead blankets, while DREAM is an helium-cooled beryllium/ceramic blanket. Both TAURO and ARIES-AT blankets are essentially formed by a SiC f /SiC box acting as a container for the lithium-lead which has the simultaneous functions of coolant, tritium breeder, neutron multiplier and, finally, tritium carrier. The DREAM blanket is characterized by small modules using pebble beds of Be as neutron multiplier material, of Li 2 O (or other lithium ceramics) as breeder material and of SiC as shielding material. The He coolant path includes a flow through the pebble beds and a porous partition wall. For each blanket, this paper describes the main design features and performances, the most recent design improvements, and the proposed manufacturing routes in order to identify specific issues and requirements for the future R and D on SiC f /SiC

To the question of definition of fissile material mass and neutron multiplication in deep sub-critical systems

International Nuclear Information System (INIS)

Dulin, V.V.

2006-01-01

A method of determination neutrons multiplication in deep sub-critical multiplying media has been developed. It is based on a modified of Rossi - alpha method. It will consist in use of integral on time (a method of the areas) from correlated parts of distribution and integral in area, independent of time a part of distribution (area of a constant background). It allows to spend the calculated analysis, using the integrated equation on time for a neutrons flux and to not use representation of point kinetic model. A calculation spatially-correlation factor the adjoint (relative the detector count rate) inhomogeneous equation is used. Its calculation takes into account fission both in multiplying media and in a spontaneous neutron source. Measurements with plutonium-steel and uranium-steel blocks, and blocks from uranium and plutonium dioxide of different enrichment are have been carried out. The measured values of neutrons multiplication in a range 1.03-1.82 will be well coordinated to results of calculations. The question on an opportunity of definition of weight of the measured blocks of fissile material is considered [ru