Thermophotovoltaic Energy Conversion in Space Nuclear Reactor Power Systems

National Research Council Canada - National Science Library

Presby, Andrew L

2004-01-01

.... This has potential benefits for space nuclear reactor power systems currently in development. The primary obstacle to space operation of thermophotovoltaic devices appears to be the low heat rejection temperatures which necessitate large radiator areas...

Progress in space nuclear reactor power systems technology development - The SP-100 program

Science.gov (United States)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

Status report on nuclear reactors for space electric power

International Nuclear Information System (INIS)

Buden, D.

1978-01-01

The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Increased questions have been raised about safety since the COSMOS 954 incident. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit where orbital lifetimes are practically indefinite, the safety considerations are negligible. The potential missions, why reactors are being considered as a prime power candidate, reactor features, and safety considerations are discussed

The role of nuclear reactors in space exploration and development

International Nuclear Information System (INIS)

Lipinski, R.J.

2000-01-01

The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of 238 Pu for power and typically generate 235 U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. One reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a new frontier for their children and their grandchildren. They pave the way for all life on earth to move out into the solar system. At one time, humans built

Selection of power plant elements for future reactor space electric power systems

International Nuclear Information System (INIS)

Buden, D.; Bennett, G.A.; Copper, K.

1979-09-01

Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected

The role of nuclear reactors in space exploration and development

Energy Technology Data Exchange (ETDEWEB)

Lipinski, R.J.

2000-07-01

The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of {sup 238}Pu for power and typically generate <1 kW of electricity. Apollo, Pioneer, Voyager, Viking, Galileo, Ulysses, and Cassini all used RTGs. Space reactors use the fission energy of {sup 235}U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. One reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a new

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

International Nuclear Information System (INIS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-01-01

A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK registered platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed

State-space model predictive control method for core power control in pressurized water reactor nuclear power stations

Energy Technology Data Exchange (ETDEWEB)

Wang, Guo Xu; Wu, Jie; Zeng, Bifan; Wu, Wangqiang; Ma, Xiao Qian [School of Electric Power, South China University of Technology, Guangzhou (China); Xu, Zhibin [Electric Power Research Institute of Guangdong Power Grid Corporation, Guangzhou (China)

2017-02-15

A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

Gaseous-fuel nuclear reactor research for multimegawatt power in space

Science.gov (United States)

Thom, K.; Schneider, R. T.; Helmick, H. H.

1977-01-01

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

Small space reactor power systems for unmanned solar system exploration missions

International Nuclear Information System (INIS)

Bloomfield, H.S.

1987-12-01

A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model

SP-100 space reactor power system readiness

International Nuclear Information System (INIS)

Josloff, A.T.; Matteo, D.N.; Bailey, H.S.

1992-01-01

This paper discusses the SP-100 Space Reactor Power System which is being developed by GE, under contract to the U.S. Department of Energy, to provide electrical power in the range of 10's to 100's of kW. The system represents an enabling technology for a wide variety of earth orbital and interplanetary science missions, nuclear electric propulsion (NEP) stages, and lunar/Mars surface power for the Space Exploration Initiative (SEI). The technology and design is now at a state of readiness to support the definition of early flight demonstration missions. Of particular importance is that SP-100 meets the demanding U.S. safety performance, reliability and life requirements. The system is scalable and flexible and can be configured to provide 10's to 100's of kWe without repeating development work and can meet DoD goals for an early, low-power demonstration flight in the 1996-1997 time frame

Nuclear space power safety and facility guidelines study

International Nuclear Information System (INIS)

Mehlman, W.F.

1995-01-01

This report addresses safety guidelines for space nuclear reactor power missions and was prepared by The Johns Hopkins University Applied Physics Laboratory (JHU/APL) under a Department of Energy grant, DE-FG01-94NE32180 dated 27 September 1994. This grant was based on a proposal submitted by the JHU/APL in response to an open-quotes Invitation for Proposals Designed to Support Federal Agencies and Commercial Interests in Meeting Special Power and Propulsion Needs for Future Space Missionsclose quotes. The United States has not launched a nuclear reactor since SNAP 10A in April 1965 although many Radioisotope Thermoelectric Generators (RTGs) have been launched. An RTG powered system is planned for launch as part of the Cassini mission to Saturn in 1997. Recently the Ballistic Missile Defense Office (BMDO) sponsored the Nuclear Electric Propulsion Space Test Program (NEPSTP) which was to demonstrate and evaluate the Russian-built TOPAZ II nuclear reactor as a power source in space. As of late 1993 the flight portion of this program was canceled but work to investigate the attributes of the reactor were continued but at a reduced level. While the future of space nuclear power systems is uncertain there are potential space missions which would require space nuclear power systems. The differences between space nuclear power systems and RTG devices are sufficient that safety and facility requirements warrant a review in the context of the unique features of a space nuclear reactor power system

Power conditioning for space nuclear reactor systems

Science.gov (United States)

Berman, Baruch

1987-01-01

This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

Alteration in reactor installation (addition of Unit 2) in Shimane Nuclear Power Station, Chugoku Electric Power Co., Inc. (inquiry)

International Nuclear Information System (INIS)

1983-01-01

An inquiry was made by the Ministry of International Trade and Industry to Nuclear Safety Commission on the addition of Unit 2 in Shimane Nuclear Power Station of The Chugoku Electric Power Co., Inc., concerning the technical capability of Chugoku Electric Power Co., Inc., and the plant safety. The NSC requested the Committee on Examination of Reactor Safety to make a deliberation on this subject. Both the technical capability and the safety of Unit 1 were already confirmed by MITI. Unit 2 to be newly added in the Shimane Nuclear Power Station is a BWR power plant with electric output of 820 MW. The examination made by MITI is described: the technical capability of Chugoku Electric Power Co., Inc., the safety of Unit 2 about its siting, reactor proper, reactor cooling system, radioactive waste management, etc. (J.P.N.)

A closed Brayton power conversion unit concept for nuclear electric propulsion for deep space missions

International Nuclear Information System (INIS)

Joyner, Claude Russell II; Fowler, Bruce; Matthews, John

2003-01-01

In space, whether in a stable satellite orbit around a planetary body or traveling as a deep space exploration craft, power is just as important as the propulsion. The need for power is especially important for in-space vehicles that use Electric Propulsion. Using nuclear power with electric propulsion has the potential to provide increased payload fractions and reduced mission times to the outer planets. One of the critical engineering and design aspects of nuclear electric propulsion at required mission optimized power levels is the mechanism that is used to convert the thermal energy of the reactor to electrical power. The use of closed Brayton cycles has been studied over the past 30 or years and shown to be the optimum approach for power requirements that range from ten to hundreds of kilowatts of power. It also has been found to be scalable to higher power levels. The Closed Brayton Cycle (CBC) engine power conversion unit (PCU) is the most flexible for a wide range of power conversion needs and uses state-of-the-art, demonstrated engineering approaches. It also is in use with many commercial power plants today. The long life requirements and need for uninterrupted operation for nuclear electric propulsion demands high reliability from a CBC engine. A CBC engine design for use with a Nuclear Electric Propulsion (NEP) system has been defined based on Pratt and Whitney's data from designing long-life turbo-machines such as the Space Shuttle turbopumps and military gas turbines and the use of proven integrated control/health management systems (EHMS). An integrated CBC and EHMS design that is focused on using low-risk and proven technologies will over come many of the life-related design issues. This paper will discuss the use of a CBC engine as the power conversion unit coupled to a gas-cooled nuclear reactor and the design trends relative to its use for powering electric thrusters in the 25 kWe to 100kWe power level

Present status of space nuclear reactor

International Nuclear Information System (INIS)

Kaneko, Yoshihiko

1996-01-01

USA and former USSR led space development, and had the experience of launching nuclear reactor satellites. In USA, the research and development of space nuclear reactor were advanced mainly by NASA, and in 1965, the nuclear reactor for power source ''SNAP-10A'' was launched and put on the orbit around the earth. Thereafter, the reactor was started up, and the verifying test at 500 We was successfully carried out. Also for developing the reactor for thermal propulsion, NERVA/ROVER project was done till 1973, and the technological basis was established. The space Exploration Initiative for sending mankind to other solar system planets than the earth is the essential point of the future projects. In former USSR, the ground experiment of the reactor for 800 We power source ''Romashka'', the development of the reactor for 10 kWe power source ''Topaz-1 and 2'', the flight of the artificial satellites, Cosmos 954 and Cosmos 1900, on which nuclear reactors were mounted, and the operation of 33 ocean-monitoring satellites ''RORSAT'' using small fast reactors were carried out. The mission of space development and the nuclear reactors as power source, the engineering of space nuclear reactors, the present status and the trend of space nuclear reactor development, and the investigation by the UN working group on the safety problem of space nuclear reactors are described. (K.I.)

Systems aspects of a space nuclear reactor power system

Science.gov (United States)

Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Bloomfield, H.; Heller, J.

1988-01-01

Various system aspects of a 300-kW nuclear reactor power system for spacecraft have been investigated. Special attention is given to the cases of a reusable OTV and a space-based radar. It is demonstrated that the stowed length of the power system is important to mission design, and that orbital storage for months to years may be needed for missions involving orbital assembly.

Review of tokamak power reactor and blanket designs in the United States

International Nuclear Information System (INIS)

Baker, C.; Brooks, J.; Ehst, D.; Gohar, Y.; Smith, D.; Sze, D.

1986-01-01

The last major conceptual design study of a tokamak power reactor in the United States was STARFIRE which was carried out in 1979-1980. Since that time US studies have concentrated on engineering test reactors, demonstration reactors, parametric systems studies, scoping studies, and studies of selected critical issues such as pulsed vs. steady-state operation and blanket requirements. During this period, there have been many advancements in tokamak physics and reactor technology, and there has also been a recognition that it is desirable to improve the tokamak concept as a commercial power reactor candidate. During 1984-1985 several organizations participated in the Tokamak Power Systems Study (TPSS) with the objective of developing ideas for improving the tokamak as a power reactor. Also, the US completed a comprehensive Blanket Comparison and Selection Study which formed the basis for further studies on improved blankets for fusion reactors

Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

International Nuclear Information System (INIS)

Koenig, D.R.; Gido, R.G.; Brandon, D.I.

1985-01-01

The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations

Solid-Core, Gas-Cooled Reactor for Space and Surface Power

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Unique features of space reactors

International Nuclear Information System (INIS)

Buden, D.

1990-01-01

This paper reports on space reactors that are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K

Development of a Robust Tri-Carbide Fueled Reactor for Multi-Megawatt Space Power and Propulsion Applications

International Nuclear Information System (INIS)

Samim Anghaie; Knight, Travis W.; Plancher, Johann; Gouw, Reza

2004-01-01

An innovative reactor core design based on advanced, mixed carbide fuels was analyzed for nuclear space power applications. Solid solution, mixed carbide fuels such as (U,Zr,Nb)c and (U,Zr, Ta)C offer great promise as an advanced high temperature fuel for space power reactors

Power unit-cargo space link in transport

Directory of Open Access Journals (Sweden)

Radmilović Zoran R.

2005-01-01

Full Text Available This paper deals with transportation technology regarding links between power unit and cargo space. These links can be divided into two groups: rigid and flexible. Rigid link, established between power unit and cargo space, is dominant in maritime and road transport (sea ships and trucks, and occasionally in transport on inland waterways (self- propelled barges. Flexible link is used in the railroad transport (systems with trailers and semi trailers, and in inland waterway transport (push - towing and pulling systems, and combinations of the systems. The main goal of this research is determination of possible link types and organization of the means of transportation.

The unique safety challenges of space reactor systems

International Nuclear Information System (INIS)

Lanes, S.J.; Marshall, A.C.

1991-01-01

Compact reactor systems can provide high levels of power for extended periods in space environments. Their relatively low mass and their ability to operate independently of their proximity to the sun make reactor power systems high desirable for many civilian and military space missions. The US Department of Energy is developing reactor system technologies to provide electrical power for space applications. In addition, reactors are now being considered to provide thermal power to a hydrogen propellant for nuclear thermal rocketry. Space reactor safety issues differ from commercial reactor issues, in some areas, because of very different operating requirements and environments. Accidents similar to those postulated for commercial reactors must be considered for space reactors during their operational phase. Safety strategies will need to be established that account for the consequences of the loss of essential power

Nuclear reactor power as applied to a space-based radar mission

Science.gov (United States)

Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Kia, T.; Nesmith, B.

1988-01-01

The SP-100 Project was established to develop and demonstrate feasibility of a space reactor power system (SRPS) at power levels of 10's of kilowatts to a megawatt. To help determine systems requirements for the SRPS, a mission and spacecraft were examined which utilize this power system for a space-based radar to observe moving objects. Aspects of the mission and spacecraft bearing on the power system were the primary objectives of this study; performance of the radar itself was not within the scope. The study was carried out by the Systems Design Audit Team of the SP-100 Project.

New generation nuclear power units of PWR type integral reactors

International Nuclear Information System (INIS)

Mitenkov, F.M.; Kurachen Kov, A.V.; Malamud, V.A.; Panov, Yu.K.; Runov, B.I.; Flerov, L.N.

1997-01-01

Design bases of new generation nuclear power units (nuclear power plants - NPP, nuclear co-generation plants - NCP, nuclear distract heating plants - NDHP), using integral type PWPS, developed in OKBM, Nizhny Novgorod and trends of design decisions optimization are considered in this report. The problems of diagnostics, servicing and repair of the integral reactor components in course of operation are discussed. The results of safety analysis, including the problems of several accident localization with postulated core melting and keeping corium in the reactor vessel and guard vessel are presented. Information on experimental substantiation of the suggested plant design decisions is presented. (author)

Summary of space nuclear reactor power systems, 1983--1992

Energy Technology Data Exchange (ETDEWEB)

Buden, D.

1993-08-11

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

Summary of space nuclear reactor power systems, 1983--1992

International Nuclear Information System (INIS)

Buden, D.

1993-01-01

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power

Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

International Nuclear Information System (INIS)

1982-01-01

A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

Reactor units for power supply to the Russian Arctic regions: Priority assessment of nuclear energy sources

Directory of Open Access Journals (Sweden)

Mel'nikov N. N.

2017-03-01

Full Text Available Under conditions of competitiveness of small nuclear power plants (SNPP and feasibility of their use to supply power to remote and inaccessible regions the competition occurs between nuclear energy sources, which is caused by a wide range of proposals for solving the problem of power supply to different consumers in the decentralized area of the Russian Arctic power complex. The paper suggests a methodological approach for expert assessment of the priority of small power reactor units based on the application of the point system. The priority types of the reactor units have been determined based on evaluation of the unit's conformity to the following criteria: the level of referentiality and readiness degree of reactor units to implementation; duration of the fuel cycle, which largely determines an autonomy level of the nuclear energy source; the possibility of creating a modular block structure of SNPP; the maximum weight of a transported single equipment for the reactor unit; service life of the main equipment. Within the proposed methodological approach the authors have performed a preliminary ranking of the reactor units according to various criteria, which allows quantitatively determining relative difference and priority of the small nuclear power plants projects aimed at energy supply to the Russian Arctic. To assess the sensitivity of the ranking results to the parameters of the point system the authors have observed the five-point and ten-point scales under variations of importance (weights of different criteria. The paper presents the results of preliminary ranking, which have allowed distinguishing the following types of the reactor units in order of their priority: ABV-6E (ABV-6M, "Uniterm" and SVBR-10 in the energy range up to 20 MW; RITM-200 (RITM-200M, KLT-40S and SVBR-100 in the energy range above 20 MW.

Space reactors - past, present, and future

International Nuclear Information System (INIS)

Buden, D.; Angelo, J.

1983-01-01

In the 1990s and beyond, advanced-design nuclear reactors could represent the prime source of both space power and propulsion. Many sophisticated military and civilian space missions of the future will require first kilowatt and then megawatt levels of power. This paper reviews key technology developments that accompanied past US space nuclear power development efforts, describes on-going programs, and then explores reactor technologies that will satisfy megawatt power level needs and beyond

Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

Science.gov (United States)

Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

2010-01-01

Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs

Energy Technology Data Exchange (ETDEWEB)

Yoder, G.L.

2005-10-03

This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.

Nuclear reactor power as applied to a space-based radar mission

Science.gov (United States)

Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Bloomfield, H.; Heller, J.

1988-01-01

A space-based radar mission and spacecraft are examined to determine system requirements for a 300 kWe space nuclear reactor power system. The spacecraft configuration and its orbit, launch vehicle, and propulsion are described. Mission profiles are addressed, and storage in assembly orbit is considered. Dynamics and attitude control and the problems of nuclear and thermal radiation are examined.

Applying design principles to fusion reactor configurations for propulsion in space

International Nuclear Information System (INIS)

Carpenter, S.A.; Deveny, M.E.; Schulze, N.R.

1993-01-01

The application of fusion power to space propulsion requires rethinking the engineering-design solution to controlled-fusion energy. Whereas the unit cost of electricity (COE) drives the engineering-design solution for utility-based fusion reactor configurations; initial mass to low earth orbit (IMLEO), specific jet power (kW(thrust)/kg(engine)), and reusability drive the engineering-design solution for successful application of fusion power to space propulsion. Three design principles (DP's) were applied to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: provide maximum direct access to space for waste radiation, operate components as passive radiators to minimize cooling-system mass, and optimize the plasma fuel, fuel mix, and temperature for best specific jet power. The three candidate terrestrial fusion reactor configurations are: the thermal barrier tandem mirror (TBTM), field reversed mirror (FRM), and levitated dipole field (LDF). The resulting three candidate space fusion propulsion systems have their IMLEO minimized and their specific jet power and reusability maximized. A preliminary rating of these configurations was performed, and it was concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System (MFPS)

Space reactors, a prospective for the future

International Nuclear Information System (INIS)

Wahlquist, E.; Voss, S.S.

1989-01-01

The power requirements for future space missions are increasing and alternate power systems will be required to meet these needs. Therefore, in the early 1980's a tri-agency space reactor program, the SP-100, was initiated that is capable of meeting the higher power requirements. To understand the current space reactor program, it is important to review it in the context of past space nuclear programs - including radioisotopes, nuclear rockets and reactors. Initial effort on these programs began in the mid-1950's. Radioisotope generators have been flown on a variety of missions and are continuing to be used. The space reactor and nuclear rocket programs were technically successful but were both terminated in 1973. The current SP-100 program builds on those earlier programs

Space nuclear reactor power plants

International Nuclear Information System (INIS)

Buden, D.; Ranken, W.A.; Koenig, D.R.

1980-01-01

Requirements for electrical and propulsion power for space are expected to increase dramatically in the 1980s. Nuclear power is probably the only source for some deep space missions and a major competitor for many orbital missions, especially those at geosynchronous orbit. Because of the potential requirements, a technology program on space nuclear power plant components has been initiated by the Department of Energy. The missions that are foreseen, the current power plant concept, the technology program plan, and early key results are described

10- to 30-kWe space power system using the uranium-zirconium hydride reactor and organic Rankine power conversion system

International Nuclear Information System (INIS)

Determan, W.R.; Bost, D.S.

1987-01-01

The UZrH reactor-ORC power system has been reviewed to determine its feasibility issues and characterize the system size, mass, and efficiency in the 10- to 30-kWe power range. The major component technologies required for this concept were reviewed to determine their technology status rating for early deployment of the system on near-term missions. Dynamic Isotope Power System (DIPS) technology is directly applicable to the UZrH reactor-ORC concept in the areas of power system reliability and survivability. The UZrH reactor-ORC concept provides a truly state-of-the-art system for use in future military and civilian space power programs. 9 references

A Sensitivity Study on the Radiation Shield of KSPR Space Reactor

Energy Technology Data Exchange (ETDEWEB)

Cerba, S.; Lee, Hyun Chul; Lim, Hong Sik; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The idea of a space reactor was realised some decades ago and since that time several research activities have been performed into this field. The US National Aeronautics and Space Administration (NASA) has been developing a small fast reactor called as fission power system (FPS) for deep space mission, where highly enriched uranium (HEU) is used as fuel. On the other hand, other researchers have also surveyed a thermal reactor concept with low enriched uranium (LEU) for space applications. One of the main concerns in terms of a space reactor is the total size and the mass of the system including the reactor itself as well as the radiation shield. Since the reactor core is a source of neutrons and gamma photons of various energies, which may cause severe damage on the electronics of the space stations, the questions related to the development of a radiation shield should be address appropriately. The proposal of a radiation shield for a small space reactor is discussed in this paper. The requirements for the radiation shield have been addressed in terms of maximal absorbed doses and neutron flounces during 10 years of operation. In this study a radiation shield design for a small space reactor was investigated. All the presented calculations were performed using the multi-purpose stochastic MCNP code with temperature dependent continuous energy ENDF/B VII.0 neutron and photon cross section libraries. The aim of this study was to design a neutron and gamma shield that can meet the requirements of 250 Gy absorbed during 10 years of reactor operation. The comparison with a fast reactor design showed that high content of {sup 238}U strongly influences the shielding mass. This phenomenon is due to the higher photon production in case of the KSPR design and therefore the use of high {sup 235}U enrichments and the operation in fast neutron spectrum may be more desirable. In case if the KSPR space reactor the best shielding performance was achieved while utilizing a multi

Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems

International Nuclear Information System (INIS)

Chow, S.

1976-01-01

A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. In the power generation mode, the plasma and propellant flows are shut off, and the driver elements supply thermal power to the power conversion system, which generates electricity for primary electric propulsion purposes

Nuclear power in space

International Nuclear Information System (INIS)

Aftergood, S.; Hafemeister, D.W.; Prilutsky, O.F.; Rodionov, S.N.; Primack, J.R.

1991-01-01

Nuclear reactors have provided energy for satellites-with nearly disastrous results. Now the US government is proposing to build nuclear-powered boosters to launch Star Wars defenses. These authors represent scientific groups that are opposed to the use of nuclear power in near space. The authors feel that the best course for space-borne reactors is to ban them from Earth orbit and use them in deep space

Space reactors - past, present, and future

International Nuclear Information System (INIS)

Buden, D.; Angelo, J.A.

1983-01-01

The successful test flights of the Space Shuttle mark the start of a new era--an era of routine manned access into cislunar space. Human technical development at the start of the next Millenium will be highlighted by the creation of Man's extraterrestrial civilization with off-planet expansion of the human resource base. In the 1990s and beyond, advanced-design nuclear reactors could represent the prime source of both space power and propulsion. Many sophisticated military and civilian space missions of the future will require first kilowatt and then megawatt levels of power. This paper reviews key technology developments that accompanied past US space nuclear power development efforts, describes on-going programs, and then explores reactor technologies that will satisfy megawatt power level needs and beyond

Accident analysis of heat pipe cooled and AMTEC conversion space reactor system

International Nuclear Information System (INIS)

Yuan, Yuan; Shan, Jianqiang; Zhang, Bin; Gou, Junli; Bo, Zhang; Lu, Tianyu; Ge, Li; Yang, Zijiang

2016-01-01

Highlights: • A transient analysis code TAPIRS for HPS has been developed. • Three typical accidents are analyzed using TAPIRS. • The reactor system has the self-stabilization ability under accident conditions. - Abstract: A space power with high power density, light weight, low cost and high reliability is of crucial importance to future exploration of deep space. Space reactor is an excellent candidate because of its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe (HP) as core cooling component, is considered as one of the most promising choices and is widely studied all over the world. This paper develops a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) based on point reactor kinetics model, lumped parameter core heat transfer model, combined HP model (self-diffusion model, flat-front startup model and network model), energy conversion model of Alkali Metal Thermal-to-Electric Conversion units (AMTEC), and HP radiator model. Three typical accidents, i.e., control drum failure, AMTEC failure and partial loss of the heat transfer area of radiator are then analyzed using TAPIRS. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. The results show the following. (1) After the failure of one set of control drums, the reactor power finally reaches a stable value after two local peaks under the temperature feedback. The fuel temperature rises rapidly, however it is still under safe limit. (2) The fuel temperature is below a safe limit under the AMTEC failure and partial loss of the heat transfer area of radiator. This demonstrates the rationality of the system design and the potential applicability of the TAPIRS code for the future engineering application of

Effects of space-dependent cross sections on core physics parameters for compact fast spectrum space power reactors

International Nuclear Information System (INIS)

Lell, R.M.; Hanan, N.A.

1987-01-01

Effects of multigroup neutron cross section generation procedures on core physics parameters for compact fast spectrum reactors have been examined. Homogeneous and space-dependent multigroup cross section sets were generated in 11 and 27 groups for a representative fast reactor core. These cross sections were used to compute various reactor physics parameters for the reference core. Coarse group structure and neglect of space-dependence in the generation procedure resulted in inaccurate computations of reactor flux and power distributions and in significant errors regarding estimates of core reactivity and control system worth. Delayed neutron fraction was insensitive to cross section treatment, and computed reactivity coefficients were only slightly sensitive. However, neutron lifetime was found to be very sensitive to cross section treatment. Deficiencies in multigroup cross sections are reflected in core nuclear design and, consequently, in system mechanical design

Nuclear reactor power for a space-based radar. SP-100 project

Science.gov (United States)

Bloomfield, Harvey; Heller, Jack; Jaffe, Leonard; Beatty, Richard; Bhandari, Pradeep; Chow, Edwin; Deininger, William; Ewell, Richard; Fujita, Toshio; Grossman, Merlin

1986-01-01

A space-based radar mission and spacecraft, using a 300 kWe nuclear reactor power system, has been examined, with emphasis on aspects affecting the power system. The radar antenna is a horizontal planar array, 32 X 64 m. The orbit is at 61 deg, 1088 km. The mass of the antenna with support structure is 42,000 kg; of the nuclear reactor power system, 8,300 kg; of the whole spacecraft about 51,000 kg, necessitating multiple launches and orbital assembly. The assembly orbit is at 57 deg, 400 km, high enough to provide the orbital lifetime needed for orbital assembly. The selected scenario uses six Shuttle launches to bring the spacecraft and a Centaur G upper-stage vehicle to assembly orbit. After assembly, the Centaur places the spacecraft in operational orbit, where it is deployed on radio command, the power system started, and the spacecraft becomes operational. Electric propulsion is an alternative and allows deployment in assembly orbit, but introduces a question of nuclear safety.

The outlook for application of powerful nuclear thermionic reactor -powered space electric jet propulsion engines

International Nuclear Information System (INIS)

Semyonov, Y.P.; Bakanov, Y.A.; Synyavsky, V.V.; Yuditsky, V.D.

1997-01-01

This paper summarizes main study results for application of powerful space electric jet propulsion unit (EJPUs) which is powered by Nuclear Thermionic Power Unit (NTPU). They are combined in Nuclear Power/Propulsion Unit (NPPU) which serves as means of spacecraft equipment power supply and spacecraft movement. Problems the paper deals with are the following: information satellites delivery and their on-orbit power supply during 10-15 years, removal of especially hazardous nuclear wastes, mining of asteroid resources and others. Evaluations on power/time/mass relationship for this type of mission are given. EJPU parameters are compatible with Russian existent or being under development launch vehicle. (author)

Gas-cooled reactor power systems for space

International Nuclear Information System (INIS)

Walter, C.E.

1987-01-01

Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

Proposal of space reactor for nuclear electric propulsion system

International Nuclear Information System (INIS)

Nishiyama, Takaaki; Nagata, Hidetaka; Nakashima, Hideki

2009-01-01

A nuclear reactor installed in spacecrafts is considered here. The nuclear reactor could stably provide an enough amount of electric power in deep space missions. Most of the nuclear reactors that have been developed up to now in the United States and the former Soviet Union have used uranium with 90% enrichment of 235 U as a fuel. On the other hand, in Japan, because the uranium that can be used is enriched to below 20%, the miniaturization of the reactor core is difficult. A Light-water nuclear reactor is an exception that could make the reactor core small. Then, the reactor core composition and characteristic are evaluated for the cases with the enrichment of the uranium fuel as 20%. We take up here Graphite reactor, Light-water reactor, and Sodium-cooled one. (author)

Systems aspects of a space nuclear reactor power system

International Nuclear Information System (INIS)

Jaffe, L.; Fujita, T.; Beatty, R.

1988-01-01

Selected systems aspects of a 300 kW nuclear reactor power system for spacecraft have been studied. The approach included examination of two candidate missions and their associated spacecraft, and a number of special topics dealing with the power system design and operation. The missions considered were a reusable orbital transfer vehicle and a space-based radar. The special topics included: Power system configuration and scaling, launch vehicle integration, operating altitude, orbital storage, start-up, thawing, control, load following, procedures in case of malfunction, restart, thermal and nuclear radiation to other portions of the spacecraft, thermal stresses between subsystems, boom and cable designs, vibration modes, attitude control, reliability, and survivability. Among the findings are that the stowed length of the power system is important to mission design and that orbital storage for months to years may be needed for missions involving orbital assembly

Estimating inhalation hazards for space nuclear power systems

International Nuclear Information System (INIS)

Hoover, M.D.; Cuddihy, R.G.; Seiler, F.Z.

1989-01-01

Minimizing inhalation hazards is a major consideration in the design, development, transportation, handling, testing, storage, launch, use, and ultimate disposition of nuclear space power systems (NSPSs). An accidental dispersion of 238 Pu is of concern for missions involving the radioisotope thermoelectric generators (RTGs) or lightweight radioisotope heater units. Materials of concern for missions involving a nuclear reactor might include other radionuclides, such as uranium, or chemically toxic materials, such as beryllium or lithium. This paper provides an overview of some of the current approaches and uncertainties associated with estimating inhalation hazards from potential NSPS accidents. The question of whether inhalation risks can be acceptable for nuclear space power systems is still open and active. The inherently low toxicity of the uranium fuel of a space nuclear reactor is a desirable feature of that option. The extensive engineering and testing that have contributed to the current generation of plutonium RTGs provide a measure of confidence that dispersion of the RTG fuel would be unlikely in an accident. The use of nuclear reactors or RTGs in space, however, requires society to assume a risk (albeit low) for dispersion of the fuel material. It can be argued that any additional risks from the use of nuclear power in space are far less than the risks we face daily

Trade study for kWe class space reactors

Science.gov (United States)

Bost, Donald S.

Recent interest by NASA and other government agencies in space reactor power systems with power levels in the 1 to 100 kWe range has prompted a review of earlier space reactor programs, as well as the ongoing SP-100 program, to identify a system that will best fulfill their needs. The candidate reactor types that were reviewed are listed. They are categorized according to the method of heat removal. The five types are: conduction cooled, heat pipe cooled, liquid metal cooled, in-core thermionic and gas cooled. The UZrH moderated reactor coupled with an organic Rankine cycle power conversion system provides an attractive system for multikilowatt, long lived missions. The reactor requires a minimum development because a similar reactor has already flown and the ORC is being developed for use in the Dynamic Isotope Power System (DIPS) and on the Space Station.

Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

Science.gov (United States)

Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

1991-01-01

The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

Multimegawatt Space Reactor Safety

International Nuclear Information System (INIS)

Stanley, M.L.

1989-01-01

The Multimegawatt (MMW) Space Reactor Project supports the Strategic Defense Initiative Office requirement to provide reliable, safe, cost-effective, electrical power in the MMW range. Specifically, power may be used for neutral particle beams, free electron lasers, electromagnetic launchers, and orbital transfer vehicles. This power plant technology may also apply to the electrical power required for other uses such as deep-space probes and planetary exploration. The Multimegawatt Space Reactor Project, the Thermionic Fuel Element Verification Program, and Centaurus Program all support the Multimegawatt Space Nuclear Power Program and form an important part of the US Department of Energy's (DOE's) space and defense power systems activities. A major objective of the MMW project is the development of a reference flight system design that provides the desired levels of public safety, health protection, and special nuclear material (SNM) protection when used during its designated missions. The safety requirements for the MMW project are a hierarchy of requirements that consist of safety requirements/regulations, a safety policy, general safety criteria, safety technical specifications, safety design specifications, and the system design. This paper describes the strategy and philosophy behind the development of the safety requirements imposed upon the MMW concept developers. The safety organization, safety policy, generic safety issues, general safety criteria, and the safety technical specifications are discussed

A preliminary feasibility study of passive in-core thermionic reactors for highly compact space nuclear power systems

International Nuclear Information System (INIS)

Parlos, A.G.; Khan, E.U.; Frymire, R.; Negron, S.; Thomas, J.K.; Peddicord, K.L.

1991-01-01

Results of a preliminary feasibility study on a new concept for a highly compact space reactor power systems are presented. Notwithstanding the preliminary nature of the present study, the results which include a new space reactor configuration and its associated technologies indicate promising avenues for the devleopment of highly compact space reactors. The calculations reported in this study include a neutronic design trade-off study using a two-dimensioinal neutron transport model, as well as a simplified one-dimensional thermal analysis of the reactor core. In arriving at the most desirable configuration, various options have been considered and analyzed, and their advantages/disadvantages have been compared. However, because of space limitation, only the most favorable reactor configuration is presented in this summary

Nuclear safety as applied to space power reactor systems

International Nuclear Information System (INIS)

Cummings, G.E.

1987-01-01

To develop a strategy for incorporating and demonstrating safety, it is necessary to enumerate the unique aspects of space power reactor systems from a safety standpoint. These features must be differentiated from terrestrial nuclear power plants so that our experience can be applied properly. Some ideas can then be developed on how safe designs can be achieved so that they are safe and perceived to be safe by the public. These ideas include operating only after achieving a stable orbit, developing an inherently safe design, ''designing'' in safety from the start and managing the system development (design) so that it is perceived safe. These and other ideas are explored further in this paper

Space nuclear power systems for extraterrestrial basing

International Nuclear Information System (INIS)

Lance, J.R.; Chi, J.W.H.

1989-01-01

Previous studies of nuclear and non-nuclear power systems for lunar bases are compared with recent studies by others. Power levels from tens of kW e for early base operation up to 2000 kW e for a self-sustaining base with a Closed Environment Life Support System (CELSS) are considered. Permanent lunar or Martian bases will require the use of multiple nuclear units connected to loads with a power transmission and distribution system analogous to earth-based electric utility systems. A methodology used for such systems is applied to the lunar base system to examine the effects of adding 100 kW e SP-100 class and/or larger nuclear units when a reliability criterion is imposed. The results show that resource and logistic burdens can be reduced by using 1000 kW e units early in the base growth scenario without compromising system reliability. Therefore, both technologies being developed in two current programs (SP-100 and NERVA Derivative Reactor (NDR) technology for space power) can be used effectively for extraterrestrial base power systems. Recent developments in NDR design that result in major reductions in reactor mass are also described. (author)

Direct Estimation of Power Distribution in Reactors for Nuclear Thermal Space Propulsion

Science.gov (United States)

Aldemir, Tunc; Miller, Don W.; Burghelea, Andrei

2004-02-01

A recently proposed constant temperature power sensor (CTPS) has the capability to directly measure the local power deposition rate in nuclear reactor cores proposed for space thermal propulsion. Such a capability reduces the uncertainties in the estimated power peaking factors and hence increases the reliability of the nuclear engine. The CTPS operation is sensitive to the changes in the local thermal conditions. A procedure is described for the automatic on-line calibration of the sensor through estimation of changes in thermal .conditions.

Power reactors in Member States. 1978 edition

International Nuclear Information System (INIS)

1978-01-01

The computer-based reactor listing gives information on reactor core characteristics and plant systems for all power reactors in operation under construction and planned. The following two tables are included to give a general picture of the overall situation: Reactor types and net electrical power; Reactor units and net electrical power by country and cumulated by year

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

Experience in start-up of the South-Ukrainian-2 power unit with the WWER-1000 reactor

International Nuclear Information System (INIS)

Belov, Yu.V.; Kazakov, V.A.; Kirilov, V.V.; Kryakvin, L.V.

1987-01-01

The volume, sequence and dates of works fulfilled during hot testing, physical and power start-ups and while bringing output of the South-Ukranian-2 power unit with the WWER-1000 reactor to the design figure are described. The works were fulfilled according to the standard schedules from October in 1984 till April in 1985. Combination of the stages and intensification of works before the physical start-up have allowed to shorten the dates by 90 days as compared to the schedule. The physical and power start-ups, including bringing reactor output to the design figure, were performed during 140 days, that permits to shorten the dates by 20 days more. The results of physical experiments carried out at the South-Ukranian-2 power unit, are in good agreement with the data obtained at the first power units of the given and Kalinin NPPs. Besides, during physical and power start-ups additional measures ensuring nuclear safety are developed

Nuclear power reactor licensing and regulation in the United States

International Nuclear Information System (INIS)

Shapar, H.K.

1979-01-01

The report is devoted to four subjects: an explanation of the origins, statutory basis and development of the present regulatory system in the United States; a description of the various actions which must be taken by a license applicant and by the Nuclear Regulatory Commission before a nuclear power plant can be constructed and placed on-line, an account of the current regulatory practices followed by the US NRC in licensing nuclear power reactors; an identification of some of the 'lessons learned' from the Three Mile Island accident and some proposed regulatory and legislative solutions. (NEA) [fr

Reactor power system deployment and startup

International Nuclear Information System (INIS)

Wetch, J.R.; Nelin, C.J.; Britt, E.J.; Klein, G.; Rasor Associates, Inc., Sunnyvale, CA; California Institute of Technology, Pasadena)

1985-01-01

This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems. 5 references

Operating US power reactors

International Nuclear Information System (INIS)

Silver, E.G.

1988-01-01

This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

The role of nuclear power and nuclear propulsion in the peaceful exploration of space

International Nuclear Information System (INIS)

2005-09-01

This publication has been produced within the framework of the IAEA's innovative reactor and fuel cycle technology development activities. It elucidates the role that peaceful space related nuclear power research and development could play in terrestrial innovative reactor and fuel cycle technology development initiatives. This review is a contribution to the Inter-Agency Meeting on Outer Space Activities, and reflects the stepped up efforts of the Scientific and Technical Subcommittee of the Committee on the Peaceful Uses of Outer Space to further strengthen cooperation between international organizations in space related activities. Apart from fostering information exchange within the United Nations organizations, this publication aims at finding new potential fields for innovative reactor and fuel cycle technology development. In assessing the status and reviewing the role of nuclear power in the peaceful exploration of space, it also aims to initiate a discussion on the potential benefits of space related nuclear power technology research and development to the development of innovative terrestrial nuclear systems

Future NASA mission applications of space nuclear power

International Nuclear Information System (INIS)

Bennett, G.L.; Mankins, J.; McConnell, D.G.; Reck, G.M.

1990-01-01

Recent studies sponsored by NASA show a continuing need for space nuclear power. A recently completed study considered missions such as a Jovian grand tour, a Uranus or Neptune orbiter and probe, and a Pluto flyby that can only be done with nuclear power. There are studies for missions beyond the outer boundaries of the solar system at distances of 100 to 1000 astronomical units. The NASA 90-day study on the space exploration initiative identified a need for nuclear reactors to power lunar surface bases and radioisotope power sources for use in lunar or Martian rovers, as well as considering options for advanced, nuclear propulsion systems for human missions to Mars

Reactor units for power supply of remote and inaccessible regions: Selection issue

Directory of Open Access Journals (Sweden)

Melnikov N.N.

2015-06-01

Full Text Available The paper briefly presents the problem aspects on power supply for the remote and inaccessible regions of Russia. Reactor units of different type and installed electric capacity have been considered in relation to the issue of power supply during mineral deposit development in the Chukotka autonomous region, Yakutia and Irkutsk region. Some preliminary assessment of the possible options for use of small nuclear power plants in various sectors of energy consumption have been carried out based on the analysis of different scenarios for economic development of the regions considered

State space modeling of reactor core in a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

2014-07-10

The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

An integrated nuclear reactor unit for a floating low capacity nuclear power plant designed for power supply in remote areas with difficult access

International Nuclear Information System (INIS)

Achkasov, A.N.; Grechko, G.I.; Gladkov, O.G.; Pavlov, V.L.; Pepa, V.N.; Shishkin, V.A.

1997-01-01

The paper describes the conceptual design of an integrated advanced safety nuclear reactor unit for a low capacity floating, NPP designed for power supply in areas which are remote with difficult access. The paper describes the major structural and lay-out components of the steam generator and reactor units with main technical characteristics. (author)

Space nuclear power and man's extraterrestrial civilization

International Nuclear Information System (INIS)

Angelo, J.J.; Buden, D.

1983-01-01

This paper examines leading space nuclear power technology candidates. Particular emphasis is given the heat-pipe reactor technology currently under development at the Los Alamos National Laboratory. This program is aimed at developing a 10-100 kWe, 7-year lifetime space nuclear power plant. As the demand for space-based power reaches megawatt levels, other nuclear reactor designs including: solid core, fluidized bed, and gaseous core, are considered

Power reactor events, May-June 1986

International Nuclear Information System (INIS)

Massaro, S.A.

1986-12-01

Power Reactor Events is a bi-monthly newsletter that compiles operating experience information about commercial nuclear power plants. This includes summaries of noteworthy events and listings and/or abstracts of USNRC and other documents that discuss safety-related or possible generic issues. It is intended to feed back some of the lessons learned from operational experience to the various plant personnel, i.e., managers, licensed reactor operators, training coordinators, and support personnel. Events at the following plants are reported: McGuire Unit 1; Susquehanna Units 1 and 2; Browns Ferry Units 1, 2, and 3; and River Bend Unit 1

Status of CEA reactor studies for a 200 kWe turboelectric Space Power System

International Nuclear Information System (INIS)

Carre, F.; Gervaise, F.; Proust, E.; Schwartz, J.P.; Tilliette, Z.; Vrillon, B.

1986-01-01

A reference design for a 200 kWe Space Nuclear Power System has been developed by the CNES and CEA Agencies of the French Government in order to assess within a first study phase running from mid 1984 to mid 1986, the key feasibility issues and the development cost of a Space Power System compatible with the version of the European launcher (ARIANE V), that will be available after 1995, and with adequate power range and lifetime performances for the missions considered at that time. The heat from a fast spectrum lithium cooled reactor is converted by a turboelectric system, selected for its technological readiness and for its advantage over thermionics and thermoelectricity, of minimizing the total mass of 100 to 300 kWe power systems, considering the available radiator area afforded by the specific ARIANE V geometrical features. A heat pipe radiator is preferred to an equivalent gas cooled system, for the increased reliability brought by the large number of independent cooling elements. The successive topics addressed in the paper, include a description of the system main components and steady state operating conditions, and the present views about the start up procedure and the reactor control

Space reactor fuels performance and development issues

International Nuclear Information System (INIS)

Wewerka, E.M.

1984-01-01

Three compact reactor concepts are now under consideration by the US Space Nuclear Power Program (the SP-100 Program) as candidates for the first 100-kWe-class space reactor. Each of these reactor designs puts unique constraints and requirements on the fuels system, and raises issues of fuel systems feasibility and performance. This paper presents a brief overview of the fuel requirements for the proposed space reactor designs, a delineation of the technical feasibility issues that each raises, and a description of the fuel systems development and testing program that has been established to address key technical issues

Fractals in Power Reactor Noise

International Nuclear Information System (INIS)

Aguilar Martinez, O.

1994-01-01

In this work the non- lineal dynamic problem of power reactor is analyzed using classic concepts of fractal analysis as: attractors, Hausdorff-Besikovics dimension, phase space, etc. A new non-linear problem is also analyzed: the discrimination of chaotic signals from random neutron noise signals and processing for diagnosis purposes. The advantages of a fractal analysis approach in the power reactor noise are commented in details

Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

Science.gov (United States)

Juhasz, Albert J.

2007-01-01

In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is

Soviet space nuclear reactor incidents - Perception versus reality

Science.gov (United States)

Bennett, Gary L.

1992-01-01

Since the Soviet Union reportedly began flying nuclear power sources in 1965 it has had four publicly known accidents involving space reactors, two publicly known accidents involving radioisotope power sources and one close call with a space reactor (Cosmos 1900). The reactor accidents, particularly Cosmos 954 and Cosmos 1402, indicated that the Soviets had adopted burnup as their reentry philosophy which is consistent with the U.S. philosophy from the 1960s and 1970s. While quantitative risk analyses have shown that the Soviet accidents have not posed a serious risk to the world's population, concerns still remain about Soviet space nuclear safety practices.

Preliminary neutronic design of spock reactor: A nuclear system for space power generation

International Nuclear Information System (INIS)

Burgio, N.; Santagata, A.; Cumo, M.; Fasano, A.; Frullini, M.

2007-01-01

Aim of this paper is to preliminary investigates the neutronic features of an upgrade of the MAUS [1] nuclear reactor whose core will be able to supply a thermoelectric converter in order to generate 30 kW of electricity for space applications. The neutronic layout of SPOCK (Space Power Core Ka) is a compact, MOX fuelled, liquid metal cooled and totally reflected fast reactor with a control system based on neutron absorption. Spock, that during the heart and launch operation must be maintained in sub-critical state, has to start up in the outer space at 40 K temperatures with the coolant in a solid state and it will reach the operating steady condition at the maximum temperature of 1300 K with the coolant in the liquid state. The main design goal is to maintains, in the operating conditions of a typical space mission, the control of the appropriate criticality margin versus temperature and coolant physical state. For this purpose, a neutronic/thermal-hydraulic calculation chain able to assists the entire design process must be set up. As preliminary recognition, MCNPX 2.5.0 and FLUENT calculations were carried out. The emerging key features of SPOCK are: an equilateral triangular mesh of 91 cylindrical UO 2 fuel rods with a Molybdenum clad ensured by two grids of the same material, cooled by liquid Sodium and contained in an AISI 316 L vessel. The core is totally wrapped by a Beryllium reflector that hosts six absorber (B 4 C) rotating control rods. The reactor shape is cylindrical (radius = 30 cm and height = 60 cm) with a total mass of 275 kg. The excess reactivity was of 5000 PCM at 1300 K. A preliminary evaluation of the control rods worth and a power spatial distribution were also discussed. Through the definition of an ideal reference K e ff value at 300 K for the actual SPOCK configuration, a sensitivity analysis on various cross sections data and material physical properties was performed for the given mission temperature range, allowing consideration on

Nuclear power in space. Use of reactors and radioactive substances as power sources in satellites and space probes; Kaernkraft i rymden. Anvaendningen av reaktorer och radioaktiva aemnen som kraftkaellor i satelliter och rymdsonder

Energy Technology Data Exchange (ETDEWEB)

Hoestbaeck, Lars

2008-11-15

Today solar panels are the most common technique to supply power to satellites. Solar panels will work as long as the power demand of the satellite is limited and the satellite can be equipped with enough panels, and kept in an orbit that allows enough sunlight to hit the panels. There are various types of space missions that do not fulfil these criteria. With nuclear power these types of missions can be powered regardless of the sunlight and as early as 1961 the first satellite with a nuclear power source was placed in orbit. Out of seventy known space missions that has made use of nuclear power, ten have had some kind of failure. In no case has the failure been associated with the nuclear technology used. This report discusses to what degree satellites with nuclear power are a source for potential radioactive contamination of Swedish territory. It is not a discussion for or against nuclear power in space. Neither is it an assessment of consequences if radioactive material from a satellite would reach the earth's surface. Historically two different kinds of Nuclear Power Sources (NPS) have been used to generate electric power in space. The first is the reactor where the energy is derived from nuclear fission of 235U and the second is the Radioisotope Thermoelectric Generator (RTG) where electricity is generated from the heat of naturally decaying radionuclides. NPS has historically only been used in space by United States and the Soviet Union (and in one failing operation Russia). Nuclear Power Sources have been used in three types of space objects: satellites, space probes and moon/Mars vehicles. USA has launched one experimental reactor into orbit, all other use of NPS by the USA has been RTG:s. The Soviet Union, in contrast, only launched a few RTG:s but nearly forty reactors. The Soviet use of NPS is less transparent than the use in USA and some data published on Soviet systems are more or less well substantiated assessments. It is likely that also future

Recent space nuclear power systems

International Nuclear Information System (INIS)

Takizuka, Takakazu; Yasuda, Hideshi; Hishida, Makoto

1991-01-01

For the advance of mankind into the space, the power sources of large output are indispensable, and it has been considered that atomic energy is promising as compared with solar energy and others. Accordingly in USA and USSR, the development of the nuclear power generation systems for space use has been carried out since considerable years ago. In this report, the general features of space nuclear reactors are shown, and by taking the system for the SP-100 project being carried out in USA as the example, the contents of the recent design regarding the safety as an important factor are discussed. Moreover, as the examples of utilizing space nuclear reactors, the concepts of the power source for the base on the moon, the sources of propulsive power for the rockets used for Mars exploration and others, the remote power transmission system by laser in the space and so on are explained. In September, 1988, the launching of a space shuttle of USA was resumed, and the Jupiter explorer 'Galileo' and the space telescope 'Hubble' were successfully launched. The space station 'Mir' of USSR has been used since February, 1986. The history of the development of the nuclear power generation systems for space use is described. (K.I.)

Power reactors in Member States. 1979 edition

International Nuclear Information System (INIS)

1979-01-01

This is the fifth issue of a periodic computer-based listing of nuclear power reactors, presenting the situation as of 1 May 1979. The basic design data for all reactors in operation, under construction, planned and shut down have been included. The following two tables are included to give a general picture of the overall situation: Table I: Reactor types and net electrical power. Table II: Reactor units and net electrical powered by country cummulated by year

Nuclear-electric power in space

International Nuclear Information System (INIS)

Truscello, V.C.; Davis, H.S.

1984-01-01

Because direct-broadcast satellites, air-traffic-control radar satellites, industrial processing on subsequent versions of the space station, and long range excursions to other planets using nuclear-electric propulsion systems, all space missions for which current power-supply systems are not sufficient. NASA and the DOE therefore have formed a joint program to develop the technology required for nuclear-reactor space power plants. After investigating potential space missions in the given range, the project will develop the technology to build such systems. High temperatures pose problems, ''hot shoes'' and ''cold shoes'', a Stirling engine dynamic system, and critical heat-transfer problems are all discussed. The nuclear reactor system for space as now envisioned is schematicized

State space model extraction of thermohydraulic systems – Part II: A linear graph approach applied to a Brayton cycle-based power conversion unit

International Nuclear Information System (INIS)

Uren, Kenneth Richard; Schoor, George van

2013-01-01

This second paper in a two part series presents the application of a developed state space model extraction methodology applied to a Brayton cycle-based PCU (power conversion unit) of a PBMR (pebble bed modular reactor). The goal is to investigate if the state space extraction methodology can cope with larger and more complex thermohydraulic systems. In Part I the state space model extraction methodology for the purpose of control was described in detail and a state space representation was extracted for a U-tube system to illustrate the concept. In this paper a 25th order nonlinear state space representation in terms of the different energy domains is extracted. This state space representation is solved and the responses of a number of important states are compared with results obtained from a PBMR PCU Flownex ® model. Flownex ® is a validated thermo fluid simulation software package. The results show that the state space model closely resembles the dynamics of the PBMR PCU. This kind of model may be used for nonlinear MIMO (multi-input, multi-output) type of control strategies. However, there is still a need for linear state space models since many control system design and analysis techniques require a linear state space model. This issue is also addressed in this paper by showing how a linear state space model can be derived from the extracted nonlinear state space model. The linearised state space model is also validated by comparing the state space model to an existing linear Simulink ® model of the PBMR PCU system. - Highlights: • State space model extraction of a pebble bed modular reactor PCU (power conversion unit). • A 25th order nonlinear time varying state space model is obtained. • Linearisation of a nonlinear state space model for use in power output control. • Non-minimum phase characteristic that is challenging in terms of control. • Models derived are useful for MIMO control strategies

Space nuclear reactors: energy gateway into the next millennium

International Nuclear Information System (INIS)

Angelo, J.A. Jr.; Buden, D.

1981-01-01

Power - reliable, abundant and economic - is the key to man's conquest of the Solar System. Space activities of the next few decades will be highlighted by the creation of the extraterrestrial phase of human civilization. Nuclear power is needed both to propel massive quantities of materials through cislunar and eventually translunar space, and to power the sophisticated satellites, space platforms, and space stations of tomorrow. To meet these anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100-kW(e) heat pipe nuclear reactor. The objectives of this program are to develop components for a space nuclear power plant capable of unattended operation for 7 to 10 years; having a reliability of greater than 0.95; and weighing less than 1910 kg. In addition, this heat pipe reactor is also compatible for launch by the US Space Transportation System

Development of Liquid-Vapor Core Reactors with MHD Generator for Space Power and Propulsion Applications

International Nuclear Information System (INIS)

Samim Anghaie

2002-01-01

Any reactor that utilizes fuel consisting of a fissile material in a gaseous state may be referred to as a gaseous core reactor (GCR). Studies on GCRs have primarily been limited to the conceptual phase, mostly due to budget cuts and program cancellations in the early 1970's. A few scientific experiments have been conducted on candidate concepts, primarily of static pressure fissile gas filling a cylindrical or spherical cavity surrounded by a moderating shell, such as beryllium, heavy water, or graphite. The main interest in this area of nuclear power generation is for space applications. The interest in space applications has developed due to the promise of significant enhancement in fuel utilization, safety, plant efficiency, special high-performance features, load-following capabilities, power conversion optimization, and other key aspects of nuclear power generation. The design of a successful GCR adapted for use in space is complicated. The fissile material studied in the pa st has been in a fluorine compound, either a tetrafluoride or a hexafluoride. Both of these molecules have an impact on the structural material used in the making of a GCR. Uranium hexafluoride as a fuel allows for a lower operating temperature, but at temperatures greater than 900K becomes essentially impossible to contain. This difficulty with the use of UF6 has caused engineers and scientists to use uranium tetrafluoride, which is a more stable molecule but has the disadvantage of requiring significantly higher operating temperatures. Gas core reactors have traditionally been studied in a steady state configuration. In this manner a fissile gas and working fluid are introduced into the core, called a cavity, that is surrounded by a reflector constructed of materials such as Be or BeO. These reactors have often been described as cavity reactors because the density of the fissile gas is low and criticality is achieved only by means of the reflector to reduce neutron leakage from the core

Some thoughts on the commercial use of reactors in space

International Nuclear Information System (INIS)

Buden, D.; Lee, J.

1986-01-01

The purpose of this paper is to illuminate the major regulatory issues associated with commercialization of space nuclear power. Currently, space reactors are government-owned and approved through the Interagency Nuclear Safety Review Panel (INSRP) and the President while commercial reactors are licensed by the Nuclear Regulatory Commission. The commercial use of reactors in space will open a new regime of regulation; that is public intervenors could enter the space licensing process for the first time. The major issue is responsibility for licensing and operations, but related considerations involve controlling special nuclear materials from aborted launches or abandoned platforms, determining final shutdown and disposal tactics, and solving new design issues such as the need for longer life of space reactor power plants

Space nuclear reactor safety

International Nuclear Information System (INIS)

Damon, D.; Temme, M.; Brown, N.

1990-01-01

Definition of safety requirements and design features of the SP-100 space reactor power system has been guided by a mission risk analysis. The analysis quantifies risk from accidental radiological consequences for a reference mission. Results show that the radiological risk from a space reactor can be made very low. The total mission risk from radiological consequences for a shuttle-launched, earth orbit SP-100 mission is estimated to be 0.05 Person-REM (expected values) based on a 1 mREM/yr de Minimus dose. Results are given for each mission phase. The safety benefits of specific design features are evaluated through risk sensitivity analyses

Gaseous fuel reactors for power systems

International Nuclear Information System (INIS)

Helmick, H.H.; Schwenk, F.C.

1978-01-01

The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems. Low power reactor experiments using uranium hexafluoride gas as fuel demonstrated performance in accordance with reactor physics predictions. The final phase of experimental activity now in progress is the fabrication and testing of a buffer gas vortex confinement system

Reactor control and protection of full scope simulator for Qinshan 300 MW Nuclear Power Unit

International Nuclear Information System (INIS)

Zhu Jinping; Sun Jiliang

1996-01-01

The control and protection simulation of Qinshan 300 MW Nuclear Power Unit, including the nuclear control, the pressurizer pressure control, the pressurizer level control, the rod control, the reactor shutdown protection and engineered safety feature etc are briefly introduced

Selected power reactor projects in Canada and the United States of America

Energy Technology Data Exchange (ETDEWEB)

NONE

1964-11-01

As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power.

Selected power reactor projects in Canada and the United States of America

International Nuclear Information System (INIS)

1964-01-01

As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power

U.S. Nuclear Power Reactor Plant Status

Data.gov (United States)

Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

Nuclear power reactors of new generation

International Nuclear Information System (INIS)

Ponomarev-Stepnoi, N.N.; Slesarev, I.S.

1988-01-01

The paper presents discussions on the following topics: fuel supply for nuclear power; expansion of the sphere of nuclear power applications, such as district heating; comparative estimates of power reactor efficiencies; safety philosophy of advanced nuclear plants, including passive protection and inherent safety concepts; nuclear power unit of enhanced safety for the new generation of nuclear power plants. The emphasis is that designers of new generation reactors face a complicated but technically solvable task of developing highly safe, efficient, and economical nuclear power sources having a wide sphere of application

Startup thaw concept for the SP-100 space reactor power system

Science.gov (United States)

Kirpich, A.; Das, A.; Choe, H.; Mcnamara, E.; Switick, D.; Bhandari, P.

1990-01-01

A thaw concept for a space reactor power system which employs lithium as a circulant for both the heat-transport and the heat-rejection fluid loops is presented. An exemplary thermal analysis for a 100-kWe (i.e., SP-100) system is performed. It is shown that the design of the thaw system requires a thorough knowledge of the various physical states of the circulant throughout the system, both spatially and temporally, and that the design has to provide adequate margins for the system to avoid a structural or thermally induced damage.

Fabrication and testing of uranium nitride fuel for space power reactors

Science.gov (United States)

Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.

1988-02-01

Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.

Hybrid Reactor designs in the United States

International Nuclear Information System (INIS)

Wolkenhauer, W.C.

1978-01-01

This paper reviews the current, active, interrelated Hybrid Reactor development programs in the United States, and offers a probable future course of action for the technology. The Department of Energy (DOE) program primarily emphasizes development of Hybrid Reactors that are optimized for proliferation resistance. The Electric Power Research Institute (EPRI) program concentrates on avenues for Hybrid Reactor commercialization. The history of electrical generation technology has been one of steady movement toward higher power densities and higher quality fuels. An apparent advantage of the Hybrid Reactor option is that it follows this trend

Compact reactor/ORC power source

International Nuclear Information System (INIS)

Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

1986-01-01

A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs

Small power reactor projects in the United States of America and Canada. Information gathered as a result of invitations from Member States

International Nuclear Information System (INIS)

1962-01-01

As part of its activities in connection with the development of nuclear power, and in response to the resolutions adopted by the General Conference, the Agency has been undertaking a continuing study of the technology and economics of small and medium sized power reactors, particularly with reference to the needs of the less-developed countries. This report summarizes the information gathered on the small power reactor projects in the United States of America and Canada, as a result of the opportunity afforded by these Member States to the Agency. It may be recalled that, at the third regular session of the General Conference, the United States Government offered to provide the Agency with relevant technical and economic data on several small power reactor projects of its Atomic Energy Commission. The Agency accepted the offer and since June 1960 it has sent one or two staff members at approximately six-monthly intervals to follow the development of nine power reactor projects in the United States which represent six different reactor systems. Last year, the Agency issued a report summarizing the information obtained through their visits and study of available published literature. The present document, which should be read in conjunction with that document, brings the information up to date and provides additional information on certain phases of the projects already discussed in the last report. Three more power reactor projects are also dealt with, namely the experimental gas-cooled reactor (EGCR), the high temperature gas-cooled reactor (HTGR) and the Hallam nuclear power facility (HNPF). Early in 1962, the Canadian Government expressed its willingness to make available to the Agency relevant information on the NPD and CANDU projects. The coverage of the NPD reactor is based upon the published information supplied by AECL of Canada and the visit by one of the staff members to the NPD site. The Agency wishes to acknowledge with thanks the co-operation extended

Radionuclide inventories for short run-time space nuclear reactor systems

International Nuclear Information System (INIS)

Coats, R.L.

1993-01-01

Space Nuclear Reactor Systems, especially those used for propulsion, often have expected operation run times much shorter than those for land-based nuclear power plants. This produces substantially different radionuclide inventories to be considered in the safety analyses of space nuclear systems. This presentation describes an analysis utilizing ORIGEN2 and DKPOWER to provide comparisons among representative land-based and space systems. These comparisons enable early, conceptual considerations of safety issues and features in the preliminary design phases of operational systems, test facilities, and operations by identifying differences between the requirements for space systems and the established practice for land-based power systems. Early indications are that separation distance is much more effective as a safety measure for space nuclear systems than for power reactors because greater decay of the radionuclide activity occurs during the time to transport the inventory a given distance. In addition, the inventories of long-lived actinides are very low for space reactor systems

Advanced Thermophotovoltaic Devices for Space Nuclear Power Systems

International Nuclear Information System (INIS)

Wernsman, Bernard; Mahorter, Robert G.; Siergiej, Richard; Link, Samuel D.; Wehrer, Rebecca J.; Belanger, Sean J.; Fourspring, Patrick; Murray, Susan; Newman, Fred; Taylor, Dan; Rahmlow, Tom

2005-01-01

Advanced thermophotovoltaic (TPV) modules capable of producing > 0.3 W/cm2 at an efficiency > 22% while operating at a converter radiator and module temperature of 1228 K and 325 K, respectively, have been made. These advanced TPV modules are projected to produce > 0.9 W/cm2 at an efficiency > 24% while operating at a converter radiator and module temperature of 1373 K and 325 K, respectively. Radioisotope and nuclear (fission) powered space systems utilizing these advanced TPV modules have been evaluated. For a 100 We radioisotope TPV system, systems utilizing as low as 2 general purpose heat source (GPHS) units are feasible, where the specific power for the 2 and 3 GPHS unit systems operating in a 200 K environment is as large as ∼ 16 We/kg and ∼ 14 We/kg, respectively. For a 100 kWe nuclear powered (as was entertained for the thermoelectric SP-100 program) TPV system, the minimum system radiator area and mass is ∼ 640 m2 and ∼ 1150 kg, respectively, for a converter radiator, system radiator and environment temperature of 1373 K, 435 K and 200 K, respectively. Also, for a converter radiator temperature of 1373 K, the converter volume and mass remains less than 0.36 m3 and 640 kg, respectively. Thus, the minimum system radiator + converter (reactor and shield not included) specific mass is ∼ 16 kg/kWe for a converter radiator, system radiator and environment temperature of 1373 K, 425 K and 200 K, respectively. Under this operating condition, the reactor thermal rating is ∼ 1110 kWt. Due to the large radiator area, the added complexity and mission risk needs to be weighed against reducing the reactor thermal rating to determine the feasibility of using TPV for space nuclear (fission) power systems

Reactor Power Meter type SG-8

Energy Technology Data Exchange (ETDEWEB)

Glowacki, S W

1981-01-01

The report describes the principle and electronic circuits of the Reactor Power Meter type SG-8. The gamma radiation caused by the activity of the reactor first cooling circuit affectes the ionization chamber being the detector of the instrument. The output detector signal direct current is converted into the frequency of electric pulses by means of the current-to-frequency converter. The output converter frequency is measured by the digital frequency meter: the number of measured digits in time unit is proportional to the reactor power.

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

Science.gov (United States)

Bowles, K. J.

1974-01-01

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

Status of CEA reactor studies for a 200 kWe turbo electric space power system

International Nuclear Information System (INIS)

Carre, F.; Gervaise, F.; Proust, E.; Schwartz, J.P.; Tilliette, Z.; Vrillon, B.

1986-01-01

The present European ARIANE space program will expand after 1995 in the development of the large ARIANE 5 launch vehicle. Considering, that the range of power needs (50 to 400 kWe) and operation times required for the space missions planned after the year 2000, are relevant to a nuclear power system, the French Centre National d'Etudes Spatiales (CNES) invited in 1983 the Commissariat a l'Energie Atomique (CEA) to undertake preliminary studies on space power systems. The purpose of the present two year phase (mid 1984-mid 1986) is to identify key technologies for a space generator within the power range of interest and to estimate the development cost of such a project to be examined for commitment in 1986. This work mainly consists in the feasibility and cost assessment of a reference 200 kWe turboelectric space generator, selected for the maturity and availability of the conversion system and for its attractive specific mass compared to thermionics and thermoelectricity, considering the available radiator area afforded by the specific ARIANE 5 geometrical features. The system is basically composed of a fast neutron spectrum lithium cooled reactor, of a Brayton conversion loop and of a heat pipe radiator

Heat transfer capability analysis of heat pipe for space reactor

International Nuclear Information System (INIS)

Li Huaqi; Jiang Xinbiao; Chen Lixin; Yang Ning; Hu Pan; Ma Tengyue; Zhang Liang

2015-01-01

To insure the safety of space reactor power system with no single point failures, the reactor heat pipes must work below its heat transfer limits, thus when some pipes fail, the reactor could still be adequately cooled by neighbor heat pipes. Methods to analyze the reactor heat pipe's heat transfer limits were presented, and that for the prevailing capillary limit analysis was improved. The calculation was made on the lithium heat pipe in core of heat pipes segmented thermoelectric module converter (HP-STMC) space reactor power system (SRPS), potassium heat pipe as radiator of HP-STMC SRPS, and sodium heat pipe in core of scalable AMTEC integrated reactor space power system (SAIRS). It is shown that the prevailing capillary limits of the reactor lithium heat pipe and sodium heat pipe is 25.21 kW and 14.69 kW, providing a design margin >19.4% and >23.6%, respectively. The sonic limit of the reactor radiator potassium heat pipe is 7.88 kW, providing a design margin >43.2%. As the result of calculation, it is concluded that the main heat transfer limit of HP-STMC SRPS lithium heat pipe and SARIS sodium heat pipe is prevailing capillary limit, but the sonic limit for HP-STMC SRPS radiator potassium heat pipe. (authors)

Primary loop simulation of the SP-100 space nuclear reactor

International Nuclear Information System (INIS)

Borges, Eduardo M.; Braz Filho, Francisco A.; Guimaraes, Lamartine N.F.

2011-01-01

Between 1983 and 1992 the SP-100 space nuclear reactor development project for electric power generation in a range of 100 to 1000 kWh was conducted in the USA. Several configurations were studied to satisfy different mission objectives and power systems. In this reactor the heat is generated in a compact core and refrigerated by liquid lithium, the primary loops flow are controlled by thermoelectric electromagnetic pumps (EMTE), and thermoelectric converters produce direct current energy. To define the system operation point for an operating nominal power, it is necessary the simulation of the thermal-hydraulic components of the space nuclear reactor. In this paper the BEMTE-3 computer code is used to EMTE pump design performance evaluation to a thermalhydraulic primary loop configuration, and comparison of the system operation points of SP-100 reactor to two thermal powers, with satisfactory results. (author)

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

A feasibility assessment of nuclear reactor power system concepts for the NASA Growth Space Station

Science.gov (United States)

Bloomfield, H. S.; Heller, J. A.

1986-01-01

A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth Space Station architecture was conducted to address a variety of installation, operational, disposition and safety issues. A previous NASA sponsored study, which showed the advantages of Space Station - attached concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide a feasibility of each combination.

A feasibility assessment of nuclear reactor power system concepts for the NASA growth Space Station

International Nuclear Information System (INIS)

Bloomfield, H.S.; Heller, J.A.

1986-01-01

A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth space station architecture was conducted to address a variety of installation, operational, disposition and safety issues. A previous NASA sponsored study, which showed the advantages of space station related concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide a feasibility of each combination

DESIGN SAFETY PROBLEMS OF NUCLEAR REACTORS IN SPACE FOR ELECTRICAL POWER

Energy Technology Data Exchange (ETDEWEB)

Pickler, D A

1963-06-15

A general treatment is presented of some of the problems in the design safety of reactors which are to be operated in space. The basic requirements of these reachigh temperatures. The usual concept of a space reactor is described briefly, and the hazards of an assumed unmanned vehicle with an enriched-U-fueled reactor are examined during its launching, orbit, and reentry. Graphs are given for the dose vs distance downwind for an excursion of 100 Mw-sec, for the activity vs time after shutdown of a reactor which has been operated for 5 yr at 100 kw(t), and for the altitude vs orbital lifetime. Apparent conflicts between the basic requirements are discussed. (D.L.C.)

Missions and planning for nuclear space power

International Nuclear Information System (INIS)

Buden, D.

1979-01-01

Requirements for electrical and propulsion power for space are expected to increase dramatically in the 1980s. Nuclear power is probably the only source for some deep space missions and a major competitor for many orbital missions, especially those at geosynchronous orbit. Because of the potential requirements, a technology program on reactor components has been initiated by the Department of Energy. The missions that are foreseen, the current reactor concept, and the technology program plan are described

A compact reactor/ORC power source

International Nuclear Information System (INIS)

Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

1986-01-01

A compact power source that combines an organic Rankine cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for componenet development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analyses combined with a finite element thermal analysis have aided in the power source design. The analysis have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high

Remote techniques for the underwater dismantling of reactor internals at the nuclear power plant Gundremmingen unit A

International Nuclear Information System (INIS)

Eickelpasch, N.; Steiner, H.; Priesmeyer, U.

1997-01-01

Unit A of the nuclear power plant in Gundremmingen (KRB A) is a boiling water reactor with an electrical power of 250 MWe. It was shut down in 1977 after eleven years of operation. The actual decommissioning started in 1983. Since then more than 5200 tons of contaminated components have been dismantled. Special cutting and handling tools were tested, developed and optimized for the purpose of working in radiation fields and under water. Due to the special design of KRB A, which uses a dual-cycle system for additional steam generation, the experience gained is transferable to pressurized water reactors. (Author)

Remote control for the underwater dismantling of reactor internals at the nuclear power plant Gundremmingen unit A

International Nuclear Information System (INIS)

Eickelpasch, N.; Steiner, H.; Priesmeyer, U.

1996-01-01

The unit A of the nuclear power plant in Gundremmingen (KRB A) is a boiling water reactor with an electrical power of 250 MW e . It was shut down in 1977 after 11 years of operation. The actual decommissioning started in 1983. Meanwhile more than 5200 tons of contaminated components have been dismantled. Special cutting and handling tools were tested, developed and optimized for the purpose of working in radiation fields and under water. Due to the special design of KRB A, using an dual cycle system for additional steam generation, the experience gained is transferable to pressurised water reactors as well. (Author)

Heat-pipe development for the SPAR space-power system

International Nuclear Information System (INIS)

Ranken, W.A.

1981-01-01

The SPAR space power system design is based on a high temperature fast spectrum nuclear reactor that furnishes heat to a thermoelectric conversion system to generate an electrical power output of 100 kW/sub (e)/. An important feature of this design is the use of alkali metal heat pipes to provide redundant, reliable, and low-loss heat transfer at high temperature. Three sets of heat pipes are used in the system. These include sodium/molybdenum heat pipes to transfer heat from the reactor core to the conversion system, potassium/niobium heat pipes to couple the conversion system to the radiator in a redundant manner, and potassium/titanium heat pipes to distribute rejected heat throughout the radiator surface. The designs of these units are discussed and fabrication methods and testing results are described. 12 figures

Impact of the use of low or medium enriched uranium on the masses of space nuclear reactor power systems

International Nuclear Information System (INIS)

1994-12-01

The design process for determining the mass increase for the substitution of low-enriched uranium (LEU) for high-enriched uranium (HEU) in space nuclear reactor systems is an optimization process which must simultaneously consider several variables. This process becomes more complex whenever the reactor core operates on an in-core thermionic power conversion, in which the fissioning of the nuclear fuel is used to directly heat thermionic emitters, with the subsequent elimination of external power conversion equipment. The increased complexity of the optimization process for this type of system is reflected in the work reported herein, where considerably more information has been developed for the moderated in-core thermionic reactors

Integrated design support systems for conceptual design of a space power reactor

International Nuclear Information System (INIS)

Shimoda, Hiroshi; Yoshikawa, Hidekazu; Takahashi, Makoto; Takeoka, Satoshi; Nagamatsu, Takashi; Ishizaki, Hiroaki

1999-01-01

In the process of conceptual design of large and complex engineering systems such as a nuclear power reactor, there must be various human works by several fields of engineers on each stage of design, analysis and evaluation. In this study, we have rearranged the design information to reduce the human workloads and have studied an efficient method to support the conceptual design works by new information technologies. For this purpose, we have developed two design support environments for conceptual design of a space power reactor as a concrete design target. When constructing an integrated design support environment, VINDS, which employs virtual reality(VR) technology, we focused on visualization of physical structure, functional organization and analysis calculation with full usage of easy perception and direct manipulation of VR. On the other hand, when constructing another asynchronous and distributed design support environment, WINDS, which employs WWW technology, we improved the support functions for cooperative design works among various fields of experts. In this paper, the basic concepts, configurations and functions of the design support environments are first described, then the future improvement is surveyed by their intercomparison. (author)

Space Nuclear Power Plant Pre-Conceptual Design Report, For Information

Energy Technology Data Exchange (ETDEWEB)

B. Levine

2006-01-27

This letter transmits, for information, the Project Prometheus Space Nuclear Power Plant (SNPP) Pre-Conceptual Design Report completed by the Naval Reactors Prime Contractor Team (NRPCT). This report documents the work pertaining to the Reactor Module, which includes integration of the space nuclear reactor with the reactor radiation shield, energy conversion, and instrumentation and control segments. This document also describes integration of the Reactor Module with the Heat Rejection segment, the Power Conditioning and Distribution subsystem (which comprise the SNPP), and the remainder of the Prometheus spaceship.

Space Fission Reactor Structural Materials: Choices Past, Present and Future

International Nuclear Information System (INIS)

Busby, Jeremy T.; Leonard, Keith J.

2007-01-01

Nuclear powered spacecraft will enable missions well beyond the capabilities of current chemical, radioisotope thermal generator and solar technologies. The use of fission reactors for space applications has been considered for over 50 years, although, structural material performance has often limited the potential performance of space reactors. Space fission reactors are an extremely harsh environment for structural materials with high temperatures, high neutron fields, potential contact with liquid metals, and the need for up to 15-20 year reliability with no inspection or preventative maintenance. Many different materials have been proposed as structural materials. While all materials meet many of the requirements for space reactor service, none satisfy all of them. However, continued development and testing may resolve these issues and provide qualified materials for space fission reactors.

Nuclear reactor descriptions for space power systems analysis

Science.gov (United States)

Mccauley, E. W.; Brown, N. J.

1972-01-01

For the small, high performance reactors required for space electric applications, adequate neutronic analysis is of crucial importance, but in terms of computational time consumed, nuclear calculations probably yield the least amount of detail for mission analysis study. It has been found possible, after generation of only a few designs of a reactor family in elaborate thermomechanical and nuclear detail to use simple curve fitting techniques to assure desired neutronic performance while still performing the thermomechanical analysis in explicit detail. The resulting speed-up in computation time permits a broad detailed examination of constraints by the mission analyst.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 2, technologies 1: Reactors, heat transport, integration issues

Science.gov (United States)

Wetch, J. R.

1988-01-01

The objectives of the Megawatt Class Nuclear Space Power System (MCNSPS) study are summarized and candidate systems and subsystems are described. Particular emphasis is given to the heat rejection system and the space reactor subsystem.

Man--machine interface issues for space nuclear power systems

International Nuclear Information System (INIS)

Nelson, W.R.; Haugset, K.

1991-01-01

The deployment of nuclear reactors in space necessitates an entirely new set of guidelines for the design of the man--machine interface (MMI) when compared to earth-based applications such as commerical nuclear power plants. Although the design objectives of earth- and space-based nuclear power systems are the same, that is, to produce electrical power, the differences in the application environments mean that the operator's role will be significantly different for space-based systems. This paper explores the issues associated with establishing the necessary MMI guidelines for space nuclear power systems. The generic human performance requirements for space-based systems are described, and the operator roles that are utilized for the operation of current and advanced earth-based reactors are briefly summarized. The development of a prototype advanced control room, the Integrated Surveillance and Control System (ISACS) at the Organization for Economic Cooperation and Development (OECD) Halden Reactor Project is introduced. Finally, preliminary ideas for the use of the ISACS system as a test bed for establishing MMI guidelines for space nuclear systems are presented

Study of advanced fission power reactor development for the United States. Volume I

International Nuclear Information System (INIS)

1976-01-01

This volume summarizes the results and conclusions of an assessment of five advanced fission power reactor concepts in the context of potential nuclear power economies developed over the time period 1975 to 2020. The study was based on the premise that the LMFBR program has been determined to be the highest priority fission reactor program and it will proceed essentially as planned. Accepting this fact, the overall objective of the study was to provide evaluations of advanced fission reactor systems for input to evaluating the levels of research and development funding for fission power. Evaluation of the reactor systems included the following categories: (1) power plant performance, (2) fuel resource utilization; (3) fuel-cycle requirements; (4) economics; (5) environmental impact; (6) risk to the public; and (7) R and D requirements to achieve commercial status. The specific major objectives of the study were twofold: (1) to parametrically assess the impact of various reactor types for various levels of power demand through the year 2020 on fissile fuel utilization, economics, and the environment, based on varying but reasonable assumptions on the rates of installation; and (2) to qualitatively assess the practicality of the advanced reactor concepts, and their research and development. The reactor concepts examined were limited to the following: advanced high-temperature, gas-cooled reactor (HTGR) systems including the thorium/U-233 fuel cycle, gas turbine, and binary cycle (BIHTGR); gas-cooled fast breeder reactor (GCFR); molten salt breeder reactor (MSBR); light water breeder reactor (LWBR); and CANDU heavy water reactor

Power unit with GT-MHR reactor plant for electricity production and district heating

International Nuclear Information System (INIS)

Kiryushin, A.L.; Kodochigov, N.G.; Kuzavkov, N.G.; Golovko, V.F.

2000-01-01

Modular helium reactor with the gas turbine (GT-MHR) is a perspective power reactor plant for the next century. The project reactor is based on experience of operation more than 50 gas-cooled reactors on carbon dioxide and helium, and also on subsequent achievements in the field of realization direct gas turbine Brayton cycle. To the beginning of 90 years, achievements in technology of gas turbines, highly effective recuperators and magnetic bearings made it possible to start development of the reactor plant project combining a safe modular gas cooled reactor and a power conversion system, realizing the highly effective Brayton cycle. The conceptual project of the commercial GT-MHR reactor plant fulfilled in 1997 by joint efforts of international firms, combines a safe modular reactor with an annular active core of prismatic fuel blocks and a power conversion system with direct gas turbine cycle. The efficiency of GT-MHR gas turbine cycle at level of about 48% makes it competitive in the electricity production market in comparison with any fossil or nuclear power stations

Cermet fuels for space power systems

International Nuclear Information System (INIS)

Barner, J.O.; Coomes, E.P.; Williford, R.E.; Neimark, L.A.

1986-01-01

A refractory-metal matrix, UN-fueled cermet is a very promising fuel candidate for a wide range of multi-megawatt space reactor systems, e.g., steady-state, flexible duty-cycle, or bimodal, single- or two-phase liquid-metal cooled reactors, or thermionic reactors. Cermet fuel is especially promising for reactor designs that require operational strategies which incorporate rapid power changes because of its anticipated capability to withstand thermal shock

System modeling and reactor design studies of the Advanced Thermionic Initiative space nuclear reactor

International Nuclear Information System (INIS)

Lee, H.H.; Abdul-Hamid, S.; Klein, A.C.

1996-01-01

In-core thermionic space reactor design concepts that operate at a nominal power output range of 20 to 50 kW(electric) are described. Details of the neutronic, thermionic, thermal hydraulics, and shielding performance are presented. Because of the strong absorption of thermal neutrons by natural tungsten and the large amount of natural tungsten within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced in-core thermionic energy conversion-based nuclear reactor systems for space applications. The results show that the driverless single-cell Advanced Thermionic Initiative (ATI) configuration, which does not have driver fuel rods, proved to be more efficient than the driven core, which has driver rods. The results also show that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one TFE works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in a 40% decrease of the efficiency of the overall system. The true axial profile is significantly different as there is a considerable amount of neutron leakage out of the top and bottom of the reactor. The analysis reveals that the axial power profile actually has a chopped cosine shape. For this axial profile, the reactor core overall efficiency for the driverless ATI reactor version is found to be 5.84% with a total electrical power of 21.92 kW(electric). By considering the true axial power profile instead of the uniform power profile, each TFE loses ∼80 W(electric)

Management of radioactive wastes at power reactor sites in India

International Nuclear Information System (INIS)

Amalraj, R.V.; Balu, K.

Indian nuclear power programme, at the present stage, is based on natural uranium fuelled heavy water moderated CANDU type reactors except for the first nuclear power station consisting of two units of enriched uranium fuelled, light water moderated, BWR type of reactors. Some of the salient aspects of radioactive waste management at power reactor sites in India are discussed. Brief reviews are presented on treatment of wastes, their disposal and environmental aspects. Indian experience in power reactor waste management is also summarised identifying some of the areas needing further work. (auth.)

Automatic power control for a pressurized water reactor

International Nuclear Information System (INIS)

Hah, Yung Joon

1994-02-01

During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

Cermet-fueled reactors for advanced space applications

International Nuclear Information System (INIS)

Cowan, C.L.; Palmer, R.S.; Taylor, I.N.; Vaidyanathan, S.; Bhattacharyya, S.K.; Barner, J.O.

1987-12-01

Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel were carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper

Applications of power beaming from space-based nuclear power stations

International Nuclear Information System (INIS)

Powell, J.R.; Botts, T.E.; Hertzberg, A.

1981-01-01

Power beaming from space-based reactor systems is examined using an advanced compact, lightweight Rotating Bed Reactor (RBR). Closed Brayton power conversion efficiencies in the range of 30 to 40% can be achieved with turbines, with reactor exit temperatures on the order of 2000 0 K and a liquid drop radiator to reject heat at temperatures of approx. 500 0 K. Higher RBR coolant temperatures (up to approx. 3000 0 K) are possible, but gains in power conversion efficiency are minimal, due to lower expander efficiency (e.g., a MHD generator). Two power beaming applications are examined - laser beaming to airplanes and microwave beaming to fixed ground receivers. Use of the RBR greatly reduces system weight and cost, as compared to solar power sources. Payback times are a few years at present prices for power and airplane fuel

Radiation protection for repairs of reactor's internals at the 2nd Unit of the Nuclear Power Plant Temelin

International Nuclear Information System (INIS)

Zapletal, P.; Konop, R.; Koc, J.; Kvasnicka, O.; Hort, M.

2011-01-01

This presentation describes the process and extent of repairs of the 2 nd unit of the Nuclear power plant Temelin during the shutdown of the reactor. All works were optimized in terms of radiation protection of workers.

Space structures, power, and power conditioning; Proceedings of the Meeting, Los Angeles, CA, Jan. 11-13, 1988

International Nuclear Information System (INIS)

Askew, R.F.

1988-01-01

Various papers on space structures, power, and power conditioning are presented. Among the topics discussed are: heterogeneous gas core reaction for space nuclear power, pulsed gas core reactor for burst power, fundamental considerations of gas core reactor systems, oscillating thermionic conversion for high-density space power, thermoelectromagnetic pumps for space nuclear power systems, lightweight electrochemical converter for space power applications, ballistic acceleration by superheated hydrogen, laser-induced current switching in gaseous discharge, electron-beam-controlled semiconductor switches, laser-controlled semiconductor closing and opening switch. Also addressed are: semiconductor-metal eutectic composites for high-power switching, optical probes for the characterization of surface breakdown, 40 kV/20 kA pseudospark switch for laser applications, insulation direction for high-power space systems, state space simulation of spacecraft power systems, structural vibration of space power station systems, minimum-time control of large space structures, novel fusion reaction for space power and propulsion, repetition rate system evaluations, cryogenic silicon photoconductive switches for high-power lasers, multilevel diamondlike carbon capacitor structure, surface breakdown of prestressed insulators, C-Mo and C-Zr alloys for space power systems, magnetic insulation for the space environment

Performance/Power Space Exploration for Binary64 Division Units

DEFF Research Database (Denmark)

Nannarelli, Alberto

2016-01-01

The digit-recurrence division algorithm is used in several high-performance processors because it provides good tradeoffs in terms of latency, area and power dissipation. In this work we develop a minimally redundant radix-8 divider for binary64 (double-precision) aiming at obtaining better energy...... efficiency in the performance-per-watt space. The results show that the radix-8 divider, when compared to radix-4 and radix-16 units, requires less energy to complete a division for high clock rates....

Power generation costs for alternate reactor fuel cycles

International Nuclear Information System (INIS)

Smolen, G.R.; Delene, J.G.

1980-09-01

The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

Low-temperature thermionics in space nuclear power systems with the safe-type fast reactor

International Nuclear Information System (INIS)

Zrodnikov, A.V.; Yarygin, V.I.; Lazarenko, G.E.; Zabudko, A.N.; Ovcharenko, M.K.; Pyshko, A.P.; Mironov, V.S.; Kuznetsov, R.V.

2007-01-01

The potentialities of the use of the low-temperature thermionic converters (TIC) with the emitter temperature ≤ 1500 K in the space nuclear power system (SNPS) with the SAFE-type (Safe Affordable Fission Engine) fast reactor proposed and developed by common efforts of American experts have been considered. The main directions of the 'SAFE-300-TEG' SNPS (300 kW(thermal)) design update by replacing the thermoelectric converters with the low-temperature high-performance thermionic converters (with the barrier index V B ≤ 1.9 eV and efficiency ≥ 10%) meant for a long-term operation (5 years at least) as the components of the SAFE-300-TIC SNPS for a Lunar base have been discussed. The concept of the SNPS with the SAFE-type fast reactor and low-temperature TICs with specific electric power of about 1.45 W/cm 2 as the components of the SAFE-300-TIC system meeting the Nasa's initial requirements to a Lunar base with the electric power demand of about 30 kW(electrical) for robotic mission has been considered. The results, involving optimization and mass-and-size estimation, show that the SAFE-300-TIC system meets the initial requirements by Nasa to the lunar base power supply. The main directions of the system update aimed at the output electric power increase up to 100 kW(electrical) have also been presented. (authors)

Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

Energy Technology Data Exchange (ETDEWEB)

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01

Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle

Space Power Program Semiannual Progress Report for period ending June 30, 1963

Energy Technology Data Exchange (ETDEWEB)

Miller, A. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

1963-10-11

This is a report of progress on the Oak Ridge National Laboratory's research and development program on nuclear power plants for electrical power production in space vehicles. The work is carried out under AEG Reactor Experiments, Fuels, and Materials, and Reactor Component programs. Research and development work is under way on the stainless steel boiling-potassium reactor and the Medium Power Reactor Experiment, boiling alkali metal heat transfer, high-temperature and refractory alloys, fuel material, and space reactor shielding, particularly in connection with SNAP 2, 8, 10, and 50. Many of these OREL efforts are directed toward the development of a specific type of power plant, but they also furnish a significant contribution of scientific and engineering information needed in other programs on advanced SNAP systems. Progress on research and development directly related to the Medium Power Reactor Experiment (MPRE) is presented mostly in Part I of this report. Progress on the MPRE will, in the future, be reported on a quarterly basis. The form of the reporting will alternate from MPRE Quarterly Progress Reports to Space Power Semiannual Progress Reports.

Specific power of liquid-metal-cooled reactors

International Nuclear Information System (INIS)

Dobranich, D.

1987-10-01

Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs

Reactor power control device

International Nuclear Information System (INIS)

Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

1998-01-01

The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

Calculation of power density with MCNP in TRIGA reactor

International Nuclear Information System (INIS)

Snoj, L.; Ravnik, M.

2006-01-01

Modern Monte Carlo codes (e.g. MCNP) allow calculation of power density distribution in 3-D geometry assuming detailed geometry without unit-cell homogenization. To normalize MCNP calculation by the steady-state thermal power of a reactor, one must use appropriate scaling factors. The description of the scaling factors is not adequately described in the MCNP manual and requires detailed knowledge of the code model. As the application of MCNP for power density calculation in TRIGA reactors has not been reported in open literature, the procedure of calculating power density with MCNP and its normalization to the power level of a reactor is described in the paper. (author)

Nuclear Reactors for Space Power, Understanding the Atom Series.

Science.gov (United States)

Corliss, William R.

The historical development of rocketry and nuclear technology includes a specific description of Systems for Nuclear Auxiliary Power (SNAP) programs. Solar cells and fuel cells are considered as alternative power supplies for space use. Construction and operation of space power plants must include considerations of the transfer of heat energy to…

Lunar electric power systems utilizing the SP-100 reactor coupled to dynamic conversion systems. Final report

International Nuclear Information System (INIS)

Harty, R.B.; Durand, R.E.

1993-03-01

An integration study was performed by Rocketdyne under contract to NASA-LeRC. The study was concerned with coupling an SP-0100 reactor to either a Brayton or Stirling power conversion system. The application was for a surface power system to supply power requirements to a lunar base. A power level of 550 kWe was selected based on the NASA Space Exploration Initiative 90-day study. Reliability studies were initially performed to determine optimum power conversion redundancy. This study resulted in selecting three operating engines and one stand-by unit. Integration design studies indicated that either the Brayton or Stirling power conversion systems could be integrated with the PS-100 reactor. The Stirling system had an integration advantage because of smaller piping size and fewer components. The Stirling engine, however, is more complex and heavier than the Brayton rotating unit, which tends to off-set the Stirling integration advantage. From a performance consideration, the Brayton had a 9 percent mass advantage, and the Stirling had a 50 percent radiator advantage

Nuclear safety. Concerns about the nuclear power reactors in Cuba

International Nuclear Information System (INIS)

Wells, Jim; Aloise, Gene; Flaherty, Thomas J.; Fitzgerald, Duane; Zavala, Mario; Hayward, Mary Alice

1992-09-01

In 1976, the Soviet Union and Cuba concluded an agreement to construct two 440-megawatt nuclear power reactors near Cienfuegos on the south central coast of Cuba, about 180 miles south of Key West, Florida. The construction of these reactors, which began around 1983, was a high priority for Cuba because of its heavy dependence on imported oil. Cuba is estimated to need an electrical generation capacity of 3,000 megawatts by the end of the decade. When completed, the first reactor unit would provide a significant percentage (estimated at over 15 percent) of Cuba's need for electricity. It is uncertain when Cuba's nuclear power reactors will become operational. On September 5, 1992, Fidel Castro announced the suspension of construction at both of Cuba's reactors because Cuba could not meet the financial terms set by the Russian government to complete the reactors. Cuban officials had initially planned to start up the first of the two nuclear reactors by the end of 1993. However, before the September 5 announcement, it was estimated that this reactor would not be operational until late 1995 or early 1996. The civil construction (such as floors and walls) of the first reactor is currently estimated to be about 90 percent to 97 percent complete, but only about 37 percent of the reactor equipment (such as pipes, pumps, and motors) has been installed. The civil construction of the second reactor is about 20 percent to 30 percent complete. No information was available about the status of equipment for the second reactor. According to former Cuban nuclear power and electrical engineers and a technician, all of whom worked at the reactor site and have recently emigrated from Cuba, Cuba's nuclear power program suffers from poor construction practices and inadequate training for future reactor operators. One former official has alleged, for example, that the first reactor containment structure, which is designed to prevent the accidental release of radioactive material into

Space nuclear reactor electric power. January 1972-May 1990 (A Bibliography from the International Aerospace Abstracts data base). Report for January 1972-May 1990

International Nuclear Information System (INIS)

1990-05-01

This bibliography contains citations concerning studies and conceptual designs of nuclear space power reactors to generate electric power for space missions. The citations cover the technology, safety aspects, and policy considerations. (This updated bibliography contains 280 citations, 172 of which are new entries to the previous edition.)

An Investigation of Power Stabilization and Space-Dependent Dynamics of a Nuclear Fluidized-Bed Reactor

International Nuclear Information System (INIS)

Pain, Christopher C.; Eaton, Matthew D.; Gomes, Jefferson L.M.A.; Oliveira, Cassiano R.E. de; Umpleby, Adrian P.; Ziver, Kemal; Ackroyd, Ron T.; Miles, Bryan; Goddard, Antony J.H.; Dam, H. van; Hagen, T.H.J.J. van der; Lathouwers, D.

2003-01-01

Previous work into the space-dependent kinetics of the conceptual nuclear fluidized bed has highlighted the sensitivity of fission power to particle movements within the bed. The work presented in this paper investigates a method of stabilizing the fission power by making it less sensitive to fuel particle movement. Steady-state neutronic calculations are performed to obtain a suitable design that is stable to radial and axial fuel particle movements in the bed. Detailed spatial/temporal simulations performed using the finite element transient criticality (FETCH) code investigate the dynamics of the new reactor design. A dual requirement of the design is that it has a moderate power output of ∼300 MW(thermal)

Nuclear Power Reactors in the World. 2016 Ed

International Nuclear Information System (INIS)

2016-01-01

Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to reactor units in IAEA Member States. This thirty-sixth edition of Reference Data Series No. 2 provides a detailed comparison of various statistics up to and including 31 December 2015. The tables and figures contain the following information: — General statistics on nuclear reactors in IAEA Member States; — Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; — Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA’s Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. Data is collected by the IAEA via designated national correspondents in Member States

Applications of nuclear-powered thermoelectric generators in space

International Nuclear Information System (INIS)

Rowe, D.M.

1991-01-01

The source of electrical power which enables information to be transmitted from the space crafts Voyager 1 and 2 back to Earth after a time period of more than a decade and at a distance of more than a billion miles is known as an RTG (radioisotope thermoelectric generator). It utilises the Seebeck effect in producing electricity from heat. In essence it consists of a large number of semiconductor thermocouples connected electrically in series and thermally in parallel. A temperature difference is maintained across the thermocouples by providing a heat source, which in the case of an RTG is a radioactive isotope, and the heat sink is space. The combination of an energy-conversion system, free of moving parts and a long-life, high energy-density heat source, provides a supply of electrical power typically in the range of tens to hundred of watts and which operates reliably over extended periods of time. An electric power source, based upon thermoelectric conversion by which utilises a nuclear reactor as a heat source, has also been deployed in space and a 100-kW system is being developed to provide electrical power to a variety of commercial and military projects including SDI. Developments in thermoelectrics that have taken place in the western world during the past 30 years are primarily due to United States interest and involvement in the exploration of space. This paper reviews US applications of nuclear-powered thermoelectric generators in space. (author)

Space reactor preliminary mechanical design

International Nuclear Information System (INIS)

Meier, K.L.

1983-01-01

An analysis was performed on the SABRE reactor space power system to determine the effect of the number and size of heat pipes on the design parameters of the nuclear subsystem. Small numbers of thin walled heat pipes were found to give a lower subsystem mass, but excessive fuel swelling resulted. The SP-100 preliminary design uses 120 heat pipes because of acceptable fuel swelling and a minimum nuclear subsystem mass of 1875 kg. Salient features of the reactor preliminary design are: individual fuel modules, ZrO 2 block core mounts, bolted collar fuel module restraints, and a BeO central plug

Guide to power reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1959-07-15

The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

Nuclear reactor power supply

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector is interposed between the protection system and the control system. This selector prevents a parameter signal of a set of signals, which differs from the other parameters signals of the set by more than twice the allowable variation of the sensors which produce the set, from passing to the control system. The selectors include a pair of signal selection units, one unit sending selected process signals to primary control channels and the other sending selected process signals to back-up control channels. Test signals are periodically impressed by a test unit on a selected pair of a selected unit and control channels. When test signals are so impressed the selected control channel is disabled from transmitting control signals to the reactor and/or its associated components. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test

Neutronics characteristics of space power reactors

International Nuclear Information System (INIS)

Little, W.; Barner, J.

1986-01-01

The objective of the paper is to describe the neutronic characteristics of a range of possible space reactor designs, and indicate the relative advantages and disadvantages of the various designs. Fuel designs to be considered are cermets (i.e., ceramic particles embedded in a metal matrix) consisting of UO 2 or Nn ceramic particles in matrices of Nb, Mo, Ta, or W. These cermet fuels are compared to a UN pin-type design. UN was selected for the reference fuel material since it has a somewhat higher density than UO 2 (i.e., 14.32 versus 10.96 gm/cc), which allows a lower minimum critical mass for both ceramic and cermet designs

Advantages and implications of U233 fueled thermionic space power energy conversion

International Nuclear Information System (INIS)

Terrell, C.W.

1992-01-01

In this paper two recent analyses are reported which demonstrate advantages of a U233 fueled thermionic fuel element (TFE) compared to 93 w/o U235, and that application (mission) has broad latitude in how space power reactor systems could or should be optimized. A reference thermionic reactor system was selected to provide the basis for the fuel comparisons. Both oxide and metal fuel forms were compared. Of special interest was to estimate the efficiencies of the four fuel forms to produce electrical power. A figure of merit (FOM) was defined which is directly proportional to the electrical average electrical power produced is proportional to the electrical power produced per unit uranium mass. In a TFE the average electrical power produced is proportional to the emitter surface area (Esa), hence the ratio Esa/Mu was selected as the FOM. Results indicate that the choice of fuel type and form leads to wide variations in critical and system masses FOM values, and system total power

Reactor power measuring device

International Nuclear Information System (INIS)

Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

1993-01-01

The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

Shielding considerations for advanced space nuclear reactor systems

International Nuclear Information System (INIS)

Angelo, J.P. Jr.; Buden, D.

1982-01-01

To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO 2 ) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications

Reference reactor module for NASA's lunar surface fission power system

International Nuclear Information System (INIS)

Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Werner, James; Qualls, Louis; Radel, Ross

2009-01-01

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO 2 -fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

SP-100 Program: space reactor system and subsystem investigations

International Nuclear Information System (INIS)

Harty, R.B.

1983-01-01

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs

Autonomous Control of Space Reactor Systems

International Nuclear Information System (INIS)

Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

2007-01-01

Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are available to perform intelligent control functions that are necessary for both normal and abnormal operational conditions

Autonomous Control of Space Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

2007-11-30

Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

Control system for a nuclear power producing unit

International Nuclear Information System (INIS)

Durrant, O.W.

1978-01-01

The invention provides in a control system for a nuclear power producing unit comprising a pressurized water reactor, a once-through steam generator provided with feedwater supply means, a turbine-generator supplied with steam from the steam generator and means maintaining a flow of pressurized water through the reactor and steam generator. The combination comprising; means generating a feed forward control signal proportional to the desired power output of the power producing unit, a second means for adjusting the reactor heat release, a third means for adjusting the rate of flow of feedwater to the steam generator, the second and third means solely responsive to and operated in parallel from the feed forward control signal whereby the reactor heat release and the rate of flow of feedwater to the steam generator are each maintained in a discrete functional relationship to the feed forward control signal

Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

International Nuclear Information System (INIS)

D. Kokkinos

2005-01-01

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

1998-01-01

To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

Space power station. Uchu hatsuden

Energy Technology Data Exchange (ETDEWEB)

Kudo, I. (Electrotechnical Laboratory, Tsukuba (Japan))

1993-02-20

A calculation tells that the amount of electric power the world will use in the future will require 100 to 500 power plants each with an output of 5-GW class. If this conception is true, it is beyond dispute that utilizing nuclear power will constitute a core of the power generation even though the geographical conditions are severe for nuclear power plants. It is also certain that power generation using clean solar energy will play important roles if power supply stability can be achieved. This paper describes plans to develop space solar power generation and space nuclear power generation that can supply power solving problems concerning geographical conditions and power supply stability. The space solar power generation is a system to arrest solar energy on a static orbit. According to a result of discussions in the U.S.A., the plan calls for solar cell sheets spread over the surface of a structure with a size of 5 km [times] 10 km [times] 0.5 km thick, and electric power obtained therefrom is transmitted to a rectenna with a size of 10 km [times] 13 km, a receiving antenna on the ground. The space nuclear power generation will be constructed similarly on a static orbit. Researches on space nuclear reactors have already begun. 10 refs., 8 figs., 1 tab.

Thermal-hydraulics for space power, propulsion, and thermal management system design

International Nuclear Information System (INIS)

Krotiuk, W.J.

1990-01-01

The present volume discusses thermal-hydraulic aspects of current space projects, Space Station thermal management systems, the thermal design of the Space Station Free-Flying Platforms, the SP-100 Space Reactor Power System, advanced multi-MW space nuclear power concepts, chemical and electric propulsion systems, and such aspects of the Space Station two-phase thermal management system as its mechanical pumped loop and its capillary pumped loop's supporting technology. Also discussed are the startup thaw concept for the SP-100 Space Reactor Power System, calculational methods and experimental data for microgravity conditions, an isothermal gas-liquid flow at reduced gravity, low-gravity flow boiling, computations of Space Shuttle high pressure cryogenic turbopump ball bearing two-phase coolant flow, and reduced-gravity condensation

Investigation of the Bilibin reactor operation in the regime of automatic power and frequency control in isolated power system

International Nuclear Information System (INIS)

Sankovskij, G.A.; Molochkov, V.I.; Dolgov, V.V.; Soldatov, G.E.; Minashin, M.E.

1981-01-01

The results of experimental investigations of the power unit operation of the Bilibin nuclear power and heating plant (BNPHP) in the regime of automatic power and frequency control in an isolated power system are presented. The BNPHP comprises four similar power units. Each unit includes a steam generating setup - the channel water-graphite reactor with tubular fuel elements with natural circulation of boiling water at all the power levels as well as a turbosetup with two heat selectors and a turbogenerator. The turbine operates on dry saturated steam (with intermediate separation) which is brought from the drum-separator of the reactor natural circulation circuit. The BNPHP operates according to the controller schedule since the start-up of the first power unit. The BNPHP unit power varifies within the 50-100% range 3-4 times per day (by the number of maxima in the schedule of the power system loadings). Two design flowsheets of the unit power control and dynamic characteristics of the system for both vatiants are considered. It is concluded that both investigated automatic control systems are seviceable and deviations of the reactor parameters within the transients are not dangerous for heat release from the core. The plant is better shielded from external mainly short-term perturbations coming from the power system when the system operates in accordance with the first variant of the flowsheet [ru

Nuclear reactor instrumentation power monitor

International Nuclear Information System (INIS)

Suzuki, Shigeru.

1989-01-01

The present invention concerns a nuclear reactor instrumentation power monitor that can be used in, for example, BWR type nuclear power plants. Signals from multi-channel detectors disposed on field units are converted respectively by LPRM signal circuits. Then, the converted signals are further converted by a multiplexer into digital signals and transmitted as serial data to a central monitor unit. The thus transmitted serial data are converted into parallel data in the signal processing section of the central monitor unit. Then, LPRM signals are taken out from each of channel detectors to conduct mathematical processing such as trip judgment or averaging. Accordingly, the field unit and the central monitor unit can be connected by way of only one data transmission cable thereby enabling to reduce the number of cables. Further, since the data are transmitted on digital form, it less undergoes effect of noises. (I.S.)

Space nuclear power: a strategy for tomorrow

International Nuclear Information System (INIS)

Buden, D.; Angelo, J. Jr.

1981-01-01

Energy: reliable, portable, abundant and low cost will be a most critical factor, perhaps the sine qua non, for the unfolding of man's permanent presence in space. Space-based nuclear power, in turn, is a key technology for developing such space platforms and the transportation systems necessary to service them. A strategy for meeting space power requirements is the development of a 100-kW(e) nuclear reactor system for high earth orbit missions, transportation from Shuttle orbits to geosynchronous orbit, and for outer planet exploration. The component technology for this nuclear power plant is now underway at the Los Alamos National Laboratory. As permanent settlements are established on the Moon and in space, multimegawatt power plants will be needed. This would involve different technology similar to terrestrial nuclear power plants

Submersion-Subcritical Safe Space (S4) reactor

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Reactor Start-up and Control Methodologies: Consideration of the Space Radiation Environment

International Nuclear Information System (INIS)

Bragg-Sitton, Shannon M.; Holloway, James Paul

2004-01-01

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable the accomplishment of ambitious space exploration missions. The natural radiation environment in space provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Initial investigation using MCNPX 2.5.b for proton transport through the SAFE-400 reactor indicates a secondary neutron net current of 1.4x107 n/s at the core-reflector interface, with an incoming current of 3.4x106 n/s due to neutrons produced in the Be reflector alone. This neutron population could provide a reliable startup source for a space reactor. Additionally, this source must be considered in developing a reliable control strategy during reactor startup, steady-state operation, and power transients. An autonomous control system is developed and analyzed for application during reactor startup, accounting for fluctuations in the radiation environment that result from changes in vehicle location (altitude, latitude, position in solar system) or due to temporal variations in the radiation field, as may occur in the case of solar flares. One proposed application of a nuclear electric propulsion vehicle is in a tour of the Jovian system, where the time required for communication to Earth is significant. Hence, it is important that a reactor control system be designed with feedback mechanisms to automatically adjust to changes in reactor temperatures, power levels, etc., maintaining nominal operation without user intervention. This paper will evaluate the potential use of secondary neutrons produced by proton interactions in the reactor vessel as a startup source for a space reactor and will present a

Small reactor power systems for manned planetary surface bases

Energy Technology Data Exchange (ETDEWEB)

Bloomfield, H.S.

1987-12-01

A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

Small reactor power systems for manned planetary surface bases

International Nuclear Information System (INIS)

Bloomfield, H.S.

1987-12-01

A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options

Historical construction costs of global nuclear power reactors

International Nuclear Information System (INIS)

Lovering, Jessica R.; Yip, Arthur; Nordhaus, Ted

2016-01-01

The existing literature on the construction costs of nuclear power reactors has focused almost exclusively on trends in construction costs in only two countries, the United States and France, and during two decades, the 1970s and 1980s. These analyses, Koomey and Hultman (2007); Grubler (2010), and Escobar-Rangel and Lévêque (2015), study only 26% of reactors built globally between 1960 and 2010, providing an incomplete picture of the economic evolution of nuclear power construction. This study curates historical reactor-specific overnight construction cost (OCC) data that broaden the scope of study substantially, covering the full cost history for 349 reactors in the US, France, Canada, West Germany, Japan, India, and South Korea, encompassing 58% of all reactors built globally. We find that trends in costs have varied significantly in magnitude and in structure by era, country, and experience. In contrast to the rapid cost escalation that characterized nuclear construction in the United States, we find evidence of much milder cost escalation in many countries, including absolute cost declines in some countries and specific eras. Our new findings suggest that there is no inherent cost escalation trend associated with nuclear technology. - Highlights: •Comprehensive analysis of nuclear power construction cost experience. •Coverage for early and recent reactors in seven countries. •International comparisons and re-evaluation of learning. •Cost trends vary by country and era; some experience cost stability or decline.

Relative power density distribution calculations of the Kori unit 1 pressurized water reactor with full-scope explicit modeling of monte carlo simulation

International Nuclear Information System (INIS)

Kim, J. O.; Kim, J. K.

1997-01-01

Relative power density distributions of the Kori unit 1 pressurized water reactor calculated by Monte Carlo modeling with the MCNP code. The Kori unit 1 core is modeled on a three-dimensional representation of the one-eighth of the reactor in-vessel component with reflective boundaries at 0 and 45 degrees. The axial core model is based on half core symmetry and is divided into four axial segments. Fission reaction density in each rod is calculated by following 100 cycles with 5,000 test neutrons in each cycle after starting with a localized neutron source and ten noncontributing settle cycles. Relative assembly power distributions are calculated from fission reaction densities of rods in assembly. After 100 cycle calculations, the system coverages to a κ value of 1.00039 ≥ 0.00084. Relative assembly power distribution is nearly the same with that of the Kori unit 1 FSAR. Applicability of the full-scope Monte Carlo simulation in the power distribution calculation is examined by the relative root mean square error of 2.159%. (author)

An overview of future sustainable nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Poullikkas, Andreas [Electricity Authority of Cyprus, P.O. Box 24506, 1399 Nicosia (Cyprus)

2013-07-01

In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA). In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will tend to have closed

Improved Nuclear Reactor and Shield Mass Model for Space Applications

Science.gov (United States)

Robb, Kevin

2004-01-01

New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

Space nuclear power systems, Part 2

International Nuclear Information System (INIS)

El-Genk, M.S.; Hoover, M.D.

1992-01-01

This volume, number two of three, contains the reviewed and edited papers were being presented at the Ninth Symposium in Albuquerque, New Mexico, 12--16 January 1992. The objective of the symposium, and hence these volumes, is to summarize the state of knowledge in the area of space nuclear power and propulsion and to provide a forum at which the most recent findings and important new developments can be presented and discussed. Topics included is this volume are: reactor and power systems control; thermionic energy conversion; space missions and power needs; key issues in nuclear and propulsion; nuclear thermal propulsion; manufacturing and processing; thermal management; space nuclear safety; and nuclear testing and production facilities

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

Operational power reactor health physics

International Nuclear Information System (INIS)

Watson, B.A.

1987-01-01

Operational Health Physics can be comprised of a multitude of organizations, both corporate and at the plant sites. The following discussion centers around Baltimore Gas and Electric's (BG and E) Calvert Cliffs Nuclear Power Plant, located in Lusby, Maryland. Calvert Cliffs is a twin Combustion Engineering 825 MWe pressurized water reactor site with Unit I having a General electric turbine-generator and Unit II having a Westinghouse turbine-generator. Having just completed each Unit's ten-year Inservice Inspection and Refueling Outge, a total of 20 reactor years operating health physics experience have been accumulated at Calvert Cliffs. Because BG and E has only one nuclear site most health physics functions are performed at the plant site. This is also true for the other BG and E nuclear related organizations, such as Engineering and Quality Assurance. Utilities with multiple plant sites have corporate health physics entity usually providing oversight to the various plant programs

System modeling for the advanced thermionic initiative single cell thermionic space nuclear reactor

International Nuclear Information System (INIS)

Lee, H.H.; Lewis, B.R.; Klein, A.C.; Pawlowski, R.A.

1993-01-01

Incore thermionic space reactor design concepts which operate in a nominal power output range of 20 to 40 kWe are described. Details of the neutronics, thermionic, shielding, and heat rejection performance are presented. Two different designs, ATI-Driven and ATI-Driverless, are considered. Comparison of the core overall performance of these two configurations are described. The comparison of these two cores includes the overall conversion efficiency, reactor mass, shield mass, and heat rejection mass. An overall system design has been developed to model the advanced incore thermionic energy conversion based nuclear reactor systems for space applications in this power range

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

Advanced Reactor Development in the United States

Energy Technology Data Exchange (ETDEWEB)

Giessing, D. F.; Griffith, J. D.; McGoff, D. J.; Rosen, Sol [U. S. Department of Energy, Texas (United States)

1990-04-15

In the United States, three technologies are employed for the new generation of advanced reactors. These technologies are Advanced Light Water Reactors (A LWRs) for the 1990s and beyond, the Modular High Temperature Gas Reactor (M HTGR) for commercial use after the turn of the century, and Liquid Metal Reactors (LWRs) to provide energy production and to convert reactor fission waste to a more manageable waste product. Each technology contributes to the energy solution. Light Water Reactors For The 1990s And Beyond--The U. S. Program The economic and national security of the United States requires a diversified energy supply base built primarily upon adequate, domestic resources that are relatively free from international pressures. Nuclear energy is a vital component of this supply and is essential to meet current and future national energy demands. It is a safe, economically continues to contribute to national energy stability, and strength. The Light Water Reactor (LWR) has been a major and successful contributor to the electrical generating needs of many nations throughout the world. It is being counted upon in the United States as a key to revitalizing nuclear energy option in the 1990s. In recent years, DOE joined with the industry to ensure the availability and future viability of the LWR option. This national program has the participation of the Nation's utility industry, the Electric Power Research Institute (EPRI), and several of the major reactor manufacturers and architect-engineers. Separate but coordinated parts of this program are managed by EPRI and DOE.

Space-reactor electric systems: subsystem technology assessment

International Nuclear Information System (INIS)

Anderson, R.V.; Bost, D.; Determan, W.R.

1983-01-01

This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified

Nuclear reactor power supply system

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector prevents a parameter signal which differs from the other parameter signals of the set by more than twice the allowable variation from passing to the control system. Test signals are periodically impressed by a test unit on a selected pair of a selection unit and control channels. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test. (author)

Brayton-Cycle Power-Conversion Unit Tested With Ion Thruster

Science.gov (United States)

Hervol, David S.

2005-01-01

Nuclear electric propulsion has been identified as an enabling technology for future NASA space science missions, such as the Jupiter Icy Moons Orbiter (JIMO) now under study. An important element of the nuclear electric propulsion spacecraft is the power conversion system, which converts the reactor heat to electrical power for use by the ion propulsion system and other spacecraft loads. The electrical integration of the power converter and ion thruster represents a key technical challenge in making nuclear electric propulsion technology possible. This technical hurdle was addressed extensively on December 1, 2003, when a closed- Brayton-cycle power-conversion unit was tested with a gridded ion thruster at the NASA Glenn Research Center. The test demonstrated end-to-end power throughput and marked the first-ever coupling of a Brayton turbo alternator and a gridded ion thruster, both of which are candidates for use on JIMO-type missions. The testing was conducted at Glenn's Vacuum Facility 6, where the Brayton unit was installed in the 3-m-diameter vacuum test port and the ion thruster was installed in the 7.6-m-diameter main chamber.

Proposal for Dual Pressurized Light Water Reactor Unit Producing 2000 MWe

International Nuclear Information System (INIS)

Kang, Kyoung Min; Noh, Sang Woo; Suh, Kune Yull

2009-01-01

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the 21 st century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well

Overview of some projects of SNPS for global space communication

International Nuclear Information System (INIS)

Ivanov, E.; Ghitaykin, V.; Ionkin, V.; Dubinin, A.; Pyshko, A.

2001-01-01

In this presentation we focused on three variants of prospective concepts of SNPS. They are intended to solve tasks of global space communication (GSC) as nearest future tasks in space. Modern concepts of the application of power technology in space believe in using an onboard source of energy for maintenance of self-transportation of the vehicle into geostationary orbit (GSO). There are three more prospective systems as follows: gas cooled nuclear reactor with hybrid thermal engine and machine power converter; nuclear reactor cooled by liquid metal and with a thermoelectric power generating system; nuclear reactor with Li cooling and a thermionic and thermoelectric power generator on board. The choice of a concept must fit strong requirements such as: space nuclear power unit is aimed to be used in a powerful mission; space power unit must be able to maintain the dual - mode regime of vehicle operation (self - transportation and long life in geosynchronous orbit [GEO]); nuclear rector of unit must be safety and it must be designed in such a way that it will ensure minimum size of the complete system; the elements of the considered technology can be used for the creation of NPPI and with other sources of heat (for example, radioisotope); the degree of technical and technological readiness of units of the thermal and power circuit of installation is estimated to be high and is defined by a number of technological developments in air, space and nuclear branches; nuclear reactor and heat transfer equipment should work in a normal mode, which can be very reliably confirmed for other high-temperature nuclear systems. Considering these concepts we practically consider one of possible strategy of developing of complex system of nuclear power engineering. It is the strategy of step-by-step development of space engineering with real application of them in commercial, scientific and other powerful missions in the nearest and deep space. As starting point of this activity is

The 1978 first in-service inspection of the reactor pressure vessel of the second unit of the Greifswald nuclear power plant

International Nuclear Information System (INIS)

Pastor, D.; Busch, R.; Hildebrandt, E.; Redlich, K.H.

1979-01-01

The reactor pressure vessel and the primary coolant circuit of the second 440-MW(e) unit of the Greifswald nuclear power plant were subjected to an in-service inspection. Extent of the inspection, development and construction of a reactor inspection container as well as the nondestructive materials testing methods used are described. Further, problems of performing the inspection, such as needs of time and personnel and radiation exposure, are considered. Finally, it is stated that the reactor pressure vessel was in safe operating state. (author)

Planning for a space infrastructure for disposal of nuclear space power systems

International Nuclear Information System (INIS)

Angelo, J. Jr.; Albert, T.E.; Lee, J.

1989-01-01

The development of safe, reliable, and compact power systems is vital to humanity's exploration, development, and, ultimately, civilization of space. Nuclear power systems appear to present to offer the only practical option of compact high-power systems. From the very beginning of US space nuclear power activities, safety has been a paramount requirement. Assurance of nuclear safety has included prelaunch ground handling operations, launch, and space operations of nuclear power sources, and more recently serious attention has been given to postoperational disposal of spent or errant nuclear reactor systems. The purpose of this paper is to describe the progress of a project to utilize the capabilities of an evolving space infrastructure for planning for disposal of space nuclear systems. Project SIREN (Search, Intercept, Retrieve, Expulsion - Nuclear) is a project that has been initiated to consider post-operational disposal options for nuclear space power systems. The key finding of Project SIREN was that although no system currently exists to affect the disposal of a nuclear space power system, the requisite technologies for such a system either exist or are planned for part of the evolving space infrastructure

Development of a Power Electronics Unit for the Space Station Plasma Contactor

Science.gov (United States)

Hamley, John A.; Hill, Gerald M.; Patterson, Michael J.; Saggio, Joseph, Jr.; Terdan, Fred; Mansell, Justin D.

1994-01-01

A hollow cathode plasma contactor has been baselined as a charge control device for the Space Station (SS) to prevent deleterious interactions of coated structural components with the ambient plasma. NASA LeRC Work Package 4 initiated the development of a plasma contactor system comprised of a Power Electronics Unit (PEU), an Expellant Management Unit (EMU), a command and data interface, and a Plasma Contactor Unit (PCU). A breadboard PEU was designed and fabricated. The breadboard PEU contains a cathode heater and discharge power supply, which were required to operate the PCU, a control and auxiliary power converter, an EMU interface, a command and telemetry interface, and a controller. The cathode heater and discharge supplies utilized a push-pull topology with a switching frequency of 20 kHz and pulse-width-modulated (PWM) control. A pulse ignition circuit derived from that used in arcjet power processors was incorporated in the discharge supply for discharge ignition. An 8088 based microcontroller was utilized in the breadboard model to provide a flexible platform for controller development with a simple command/data interface incorporating a direct connection to SS Mulitplexer/Demultiplexer (MDM) analog and digital I/O cards. Incorporating this in the flight model would eliminate the hardware and software overhead associated with a 1553 serial interface. The PEU autonomously operated the plasma contactor based on command inputs and was successfully integrated with a prototype plasma contactor unit demonstrating reliable ignition of the discharge and steady-state operation.

Thulium heat sources for space power applications

International Nuclear Information System (INIS)

Alderman, C.J.

1992-05-01

Reliable power supplies for use in transportation and remote systems will be an important part of space exploration terrestrial activities. A potential power source is available in the rare earth metal, thulium. Fuel sources can be produced by activating Tm-169 targets in the space station reactor. The resulting Tm-170 heat sources can be used in thermoelectric generators to power instrumentation and telecommunications located at remote sites such as weather stations. As the heat source in a dynamic Sterling or Brayton cycle system, the heat source can provide a lightweight power source for rovers or other terrestrial transportation systems

Space reactor/organic Rankine conversion - A near-term state-of-the-art solution

Science.gov (United States)

Niggemann, R. E.; Lacey, D.

The use of demonstrated reactor technology with organic Rankine cycle (ORC) power conversion can provide a low cost, minimal risk approach to reactor-powered electrical generation systems in the near term. Several reactor technologies, including zirconium hydride, EBR-II and LMFBR, have demonstrated long life and suitability for space application at the operating temperature required by an efficient ORC engine. While this approach would not replace the high temperature space reactor systems presently under development, it could be available in a nearer time frame at a low and predictable cost, allowing some missions requiring high power levels to be flown prior to the availability of advanced systems with lower specific mass. Although this system has relatively high efficiency, the heat rejection temperature is low, requiring a large radiator on the order of 3.4 sq m/kWe. Therefore, a deployable heat pipe radiator configuration will be required.

Green Applications for Space Power

Data.gov (United States)

National Aeronautics and Space Administration — Spacecraft propulsion and power for many decades has relied on Hydrazine monopropellant technology for auxiliary power units (APU), orbital circularization, orbit...

Fuzzy power control algorithm for a pressurized water reactor

International Nuclear Information System (INIS)

Hah, Y.J.; Lee, B.W.

1994-01-01

A fuzzy power control algorithm is presented for automatic reactor power control in a pressurized water reactor (PWR). Automatic power shape control is complicated by the use of control rods with a conventional proportional-integral-differential controller because it is highly coupled with reactivity compensation. Thus, manual shape controls are usually employed even for the limited capability needed for load-following operations including frequency control. In an attempt to achieve automatic power shape control without any design modifications to the core, a fuzzy power control algorithm is proposed. For the fuzzy control, the rule base is formulated based on a multiple-input multiple-output system. The minimum operation rule and the center of area method are implemented for the development of the fuzzy algorithm. The fuzzy power control algorithm has been applied to Yonggwang Nuclear Unit 3. The simulation results show that the fuzzy control can be adapted as a practical control strategy for automatic reactor power control of PWRs during the load-following operations

Reference reactor module for NASA's lunar surface fission power system

Energy Technology Data Exchange (ETDEWEB)

Poston, David I [Los Alamos National Laboratory; Kapernick, Richard J [Los Alamos National Laboratory; Dixon, David D [Los Alamos National Laboratory; Werner, James [INL; Qualls, Louis [ORNL; Radel, Ross [SNL

2009-01-01

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

United States Domestic Research Reactor Infrastructure - TRIGA Reactor Fuel Support

International Nuclear Information System (INIS)

Morrell, Douglas

2008-01-01

The purpose of the United State Domestic Research Reactor Infrastructure Program is to provide fresh nuclear reactor fuel to United States universities at no, or low, cost to the university. The title of the fuel remains with the United States government and when universities are finished with the fuel, the fuel is returned to the United States government. The program is funded by the United States Department of Energy - Nuclear Energy division, managed by Department of Energy - Idaho Field Office, and contracted to the Idaho National Laboratory's Management and Operations Contractor - Battelle Energy Alliance. Program has been at Idaho since 1977 and INL subcontracts with 26 United States domestic reactor facilities (13 TRIGA facilities, 9 plate fuel facilities, 2 AGN facilities, 1 Pulstar fuel facility, 1 Critical facility). University has not shipped fuel since 1968 and as such, we have no present procedures for shipping spent fuel. In addition: floor loading rate is unknown, many interferences must be removed to allow direct access to the reactor tank, floor space in the reactor cell is very limited, pavement ends inside our fence; some of the surface is not finished. The whole approach is narrow, curving and downhill. A truck large enough to transport the cask cannot pull into the lot and then back out (nearly impossible / refused by drivers); a large capacity (100 ton), long boom crane would have to be used due to loading dock obstructions. Access to the entrance door is on a sidewalk. The campus uses it as a road for construction equipment, deliveries and security response. Large trees are on both sides of sidewalk. Spent fuel shipments have never been done, no procedures approved or in place, no approved casks, no accident or safety analysis for spent fuel loading. Any cask assembly used in this facility will have to be removed from one crane, moved on the floor and then attached to another crane to get from the staging area to the reactor room. Reactor

Space dependence of reactivity parameters on reactor dynamic perturbation measurements

International Nuclear Information System (INIS)

Maletti, R.; Ziegenbein, D.

1985-01-01

Practical application of reactor-dynamic perturbation measurements for on-power determination of differential reactivity weight of control rods and power coefficients of reactivity has shown a significant dependence of parameters on the position of outcore detectors. The space dependence of neutron flux signal in the core of a VVER-440-type reactor was measured by means of 60 self-powered neutron detectors. The greatest neutron flux alterations are located close to moved control rods and in height of the perturbation position. By means of computations, detector positions can be found in the core in which the one-point model is almost valid. (author)

The working lifetime of nuclear power plants and new types of power reactors

International Nuclear Information System (INIS)

Bataille, Ch.; Birraux, C.

2003-01-01

The report on the working lifetime of nuclear power plants and new reactor types, by Mr Christian Bataille, deputy for the Nord, and Mr Claude Birraux, deputy for Haute-Savoie as well as President of the Office, supplements the studies carried out by the Parliamentary Office on the Safety of Nuclear Installations and Radioactive Wastes: it examines the remaining working life of the EDF nuclear power plants and the current status of projects that might, if circumstances were right, replace the reactors at present in service. The report investigates the different physical and other factors that influence the ageing of nuclear power plants and tackles the question of whether the design life of 40 years could be exceeded in practice. The whole issue of French nuclear power plant is put in perspective and compared with the situation of nuclear plants in Finland, Sweden, Germany and the United States, from the technical and regulatory standpoints. Believing that any attempt to optimise the working lifetime of the power plants currently in service must be accompanied by simultaneous moves aimed at their replacement, Messrs. Christian Bataille and Claude Birraux go on to review in detail the various light water reactor projects being proposed around the world for completion by 2015, as developments of existing models, in particular the EPR reactor of Framatome ANP, characterised by its competitiveness. They suggest that a first such reactor should be built as quickly as possible. Describing the other nuclear systems being investigated by research organisations not only in France but also in the United States and Sweden, Mrs. Christian Bataille and Claude Birraux review the objectives of these and the circumstances in which they might be developed, which would be unlikely to be before 2035 in view of the technological problems to be overcome and the industrial demonstration plants that would be needed

Modeling and analysis of the disk MHD generator component of a gas core reactor/MHD Rankine cycle space power system

International Nuclear Information System (INIS)

Welch, G.E.; Dugan, E.T.; Lear, W.E. Jr.; Appelbaum, J.G.

1990-01-01

A gas core nuclear reactor (GCR)/disk magnetohydrodynamic (MHD) generator direct closed Rankine space power system concept is described. The GCR/disk MHD generator marriage facilitates efficient high electric power density system performance at relatively high operating temperatures. The system concept promises high specific power levels, on the order of 1 kW e /kg. An overview of the disk MHD generator component magnetofluiddynamic and plasma physics theoretical modeling is provided. Results from a parametric design analysis of the disk MHD generator are presented and discussed

Control unit of a nuclear reactor

International Nuclear Information System (INIS)

Desfontaines, Guy; Le Helloco, Michel.

1981-01-01

Control unit comprising multiple leak-tight vessels, in communication with the inside of the reactor vessel, extending this vessel above its cover, in the vertical direction and each one enclosing a mechanism for moving a cluster of material absorbing the neutrons in the reactor core, actuated by a motor. This control unit is of reduced height above the vessel cover and provides efficient protection of the leak tight containments and the mechanisms in the event of earthquakes, easy removal and refitting of the vessel cover, good ventilation of the power devices of the mechanisms without the use of a complex ventilation system, efficient thermal insulation of the leak-tight containments assembly, as well as easy access to the motors and mechanism located in the leak-tight containment for carrying out any maintenance and repairs that might be required [fr

Reactor power control device

International Nuclear Information System (INIS)

Doi, Kazuyori.

1981-01-01

Purpose: To automatically control the BWR type reactor power by simple and short-time searching the load pattern nearest to the required pattern at a nuclear power plant side. Constitution: The reactor power is automatically regulated by periodical modifying of coefficients fitting to a reactor core model, according as a required load pattern. When a load requirement pattern is given, a simulator estimates the total power change and the axial power distribution change from a xenon density change output calculated by a xenon dynamic characteristic estimating device, and a load pattern capable of being realized is searched. The amount to be recirculated is controlled on the basis of the load patteren thus searched, and the operation of the BWR type reactor is automatically controlled at the side of the nuclear power plant. (Kamimura, M.)

Nuclear Power: Outlook for New U.S. Reactors

National Research Council Canada - National Science Library

Parker, Larry; Holt, Mark

2007-01-01

Nearly three decades after the most recent order was placed for a new nuclear power plant in the United States, several utilities are now expressing interest in building a total of up to 30 new reactors...

The decommissioning of commercial magnox gas cooled reactor power stations in the United Kingdom

International Nuclear Information System (INIS)

Holt, G.

1998-01-01

There are nine commercial Magnox gas-cooled reactor power stations in the United Kingdom. Three of these stations have been shutdown and are being decommissioning, and plans have also been prepared for the eventual decommissioning of the remaining operational stations. The preferred strategy for the decommissioning of the Magnox power stations has been identified as 'Safestore' in which the decommissioning activities are carried out in a number of steps separated by quiescent periods of care and maintenance. The final clearance of the site could be deferred for up to 135 years following station shutdown so as to obtain maximum benefit from radioactive decay. The first step in the decommissioning strategy is to defuel the reactors and transport all spent and new fuel off the site. This work has been completed at all three shutdown stations. Decommissioning work is continuing on the three sites and has involved activities such as dismantling, decontamination, recycling and disposal of some plant and structures, and the preparation of others for retention on the site for a period of care and maintenance. Significant experience has been gained in the practical application of decommissioning, with successful technologies and processes being identified for a wide range of activities. For example, large and small metallic and concrete structures, some with complex geometries, have been successfully decontaminated. Also, the reactors have been prepared for a long period of care and maintenance, with instrumentation and sampling systems having been installed to monitor their continuing integrity. All of this work has been done under careful safety, technical, and financial control. (author)

SP-100 space nuclear power system

International Nuclear Information System (INIS)

Given, R.W.; Morgan, R.E.; Chi, J.W.H.; Westinghouse Electric Corp., Madison, PA)

1984-01-01

A baseline design concept for a 100 kWe nuclear reactor space power system is described. The concept was developed under contract from JPL as part of a joint program of the DOE, DOD, and NASA. The major technical and safety constraints influencing the selection of reactor operating parameters are discussed. A lithium-cooled compact fast reactor was selected as the best candidate system. The material selected for the thermoelectric conversion system was silicon germanium (SiGe) with gallium phosphide doping. Attention is given to the improved safety of the seven in-core control rod configuration

Technical evaluation of the proposed deletion of a reactor trip on a turbine trip below 50-percent power for the Beaver Valley nuclear power plant, Unit 1

International Nuclear Information System (INIS)

Reeves, W.E.

1979-12-01

This report documents the technical evaluation of the Duquesne Light Company's proposed license amendment for the deletion of a reactor trip on a turbine trip below 50% power for the Beaver Valley nuclear power plant, Unit 1. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Program being conducted for the US Nuclear Regulatory Commission by Lawrence Livermore Laboratory

Compatibility of Space Nuclear Power Plant Materials in an Inert He/Xe Working Gas Containing Reactive Impurities

International Nuclear Information System (INIS)

MM Hall

2006-01-01

A major materials selection and qualification issue identified in the Space Materials Plan is the potential for creating materials compatibility problems by combining dissimilar reactor core, Brayton Unit and other power conversion plant materials in a recirculating, inert He/Xe gas loop containing reactive impurity gases. Reported here are results of equilibrium thermochemical analyses that address the compatibility of space nuclear power plant (SNPP) materials in high temperature impure He gas environments. These studies provide early information regarding the constraints that exist for SNPP materials selection and provide guidance for establishing test objectives and environments for SNPP materials qualification testing

Power control system in BWR type reactors

International Nuclear Information System (INIS)

Nishizawa, Yasuo.

1980-01-01

Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

Space Station Freedom power - A reliability, availability, and maintainability assessment of the proposed Space Station Freedom electric power system

Science.gov (United States)

Turnquist, S. R.; Twombly, M.; Hoffman, D.

1989-01-01

A preliminary reliability, availability, and maintainability (RAM) analysis of the proposed Space Station Freedom electric power system (EPS) was performed using the unit reliability, availability, and maintainability (UNIRAM) analysis methodology. Orbital replacement units (ORUs) having the most significant impact on EPS availability measures were identified. Also, the sensitivity of the EPS to variations in ORU RAM data was evaluated for each ORU. Estimates were made of average EPS power output levels and availability of power to the core area of the space station. The results of assessments of the availability of EPS power and power to load distribution points in the space stations are given. Some highlights of continuing studies being performed to understand EPS availability considerations are presented.

Atomic power in space: A history

International Nuclear Information System (INIS)

1987-03-01

''Atomic Power in Space,'' a history of the Space Isotope Power Program of the United States, covers the period from the program's inception in the mid-1950s through 1982. Written in non-technical language, the history is addressed to both the general public and those more specialized in nuclear and space technologies. 19 figs., 3 tabs

Gas-core reactor power transient analysis. Final report

International Nuclear Information System (INIS)

Kascak, A.F.

1972-01-01

The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of the study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process. (auth)

Comparison of Ontario Hydro's performance with world power reactors - 1981

International Nuclear Information System (INIS)

Dumka, B.R.

1982-04-01

The performance of Ontario Hydro's CANDU reactors in 1981 is compared with that of 123 world nuclear power reactors rated at 500 MW(e) or greater. The report is based on data extracted from publications, as well as correspondence with a number of utilities. The basis used is the gross capacity factor, which is defined as gross unit generation divided by the perfect gross output for the period of interest. The lowest of the published turbine and generator design ratings is used to determine the perfect gross output, unless the unit has been proven capable of consistently exceeding this value. The first six reactors in the rankings were CANDU reactors operated by Ontario Hydro

Monitoring device for the power distribution within a nuclear reactor core

International Nuclear Information System (INIS)

Tanzawa, Tomio; Kumanomido, Hironori; Toyoshi, Isamu.

1986-01-01

Purpose: To provide a monitoring device for the power distribution in the reactor core that calculates the power distribution based on the measurement by instruments disposed within the reactor core of BWR type reactors. Constitution: The power distribution monitoring device in a reactor core comprises a signal correcting device, a signal normalizing device and a power distribution calculating device, in which the power distribution calculating device is constituted with an average power calculating device for four fuel assemblies and an average power calculating device for fuel assemblies. Gamma-ray signals corrected by the signal correcting device and signals from neutron detectors are inputted to the signal normalizing device, both of which are calibrated to determine the axial gamma-ray signal distribution in the central water gap region with the four fuel assemblies being as the unit. The average power from the four fuel assemblies are inputted to the fuel assembly average power calculating device to allocate to each of the fuel assembly average power thereby attaining the purpose. Further, thermal restriction values are calculated thereby enabling to secure the fuel integrity. (Kamimura, M.)

Radiation risk from the nuclear power installation of space vehicle in case of reentry to the atmosphere

International Nuclear Information System (INIS)

Mikheenko, S.G.

1994-01-01

Main directions of space using of nuclear power are considered. Nuclear energy has found many applications in space projects. The first application is the use of nuclear energy for the production of electricity in space and the second main application is the use of nuclear power for propulsion purposes in space flight. History of usage nuclear power systems in space technic is shown. Today there are 54 satellites with NPS in space near the Earth. The main principle of radical solution of the problem of radiation safety is based on the accommodation of space objects with nuclear units in orbits, such that the ballistic lifetime is greater than the time necessary for complete decay of the accumulated radioactivity. Radiation safety on various stages of space nuclear systems exploitation is discussed. If Main System Ensuring Radiation Safety is failed, it must operates Reserved System Ensuring Radiation Safety. Concrete development of a booster system for nuclear unit and a system for the reactor destruction in order to ensure aerodynamic destruction of fuel has been realized in satellite of 'Cosmos' series. The investigations on reserved system ensuring radiation safety in Moscow Physical - Engineering Institute are discussed. The results show that we can in principle ensure the radiation safety in accordance to ICRP recommendations. (author)

Calibration of RB reactor power

International Nuclear Information System (INIS)

Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

1976-09-01

The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

OHara J. M.; Higgins, J.; DAgostino, A.

2012-01-17

The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a single operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.

Different types of power reactors and provenness

International Nuclear Information System (INIS)

Goodman, E.I.

1977-01-01

The lecture guides the potential buyer in the selection of a reactor type. Recommended criteria regarding provenness, licensability, and contractual arrangements are defined and discussed. Tabular data summarizing operating experience and commercial availability of units are presented and discussed. The status of small and medium power reactors which are of interest to many developing countries is presented. It is stressed that each prospective buyer will have to establish his own criteria based on specific conditions which will be applied to reactor selection. In all cases it will be found that selection, either pre-selection of bidders or final selection of supplier, will be a fairly complex evaluation. (orig.) [de

Atomic power in space: A history

Energy Technology Data Exchange (ETDEWEB)

1987-03-01

''Atomic Power in Space,'' a history of the Space Isotope Power Program of the United States, covers the period from the program's inception in the mid-1950s through 1982. Written in non-technical language, the history is addressed to both the general public and those more specialized in nuclear and space technologies. 19 figs., 3 tabs.

Fuzzy algorithm for an automatic reactor power control in a PWR

International Nuclear Information System (INIS)

Hah, Yung Joon; Song, In Ho; Yu, Sung Sik; Choi, Jung In; Lee, Byong Whi

1994-01-01

A fuzzy algorithm is presented for automatic reactor power control in a pressurized water reactor. Automatic power shape control is complicated by the use of control rods because it is highly coupled with reactivity compensation. Thus, manual shape controls are usually employed even for the limited capability for the load - follow operation including frequency control. In an attempt to achieve automatic power shape control without any design modification of the core, a fuzzy power control algorithm is proposed. For the fuzzy control, the rule base is formulated based on a multi - input multi - output system. The minimum operation rule and the center of area method are implemented for the development of the fuzzy algorithm. The fuzzy power control algorithm has been applied to the Yonggwang Nuclear Unit 3. The simulation results show that the fuzzy control can be adapted as a practical control strategy for automatic reactor power control of the pressurized water reactor during the load - follow operation

Reactor power monitoring device

International Nuclear Information System (INIS)

Dogen, Ayumi; Ozawa, Michihiro.

1983-01-01

Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

Study of space reactors for exploration missions

Energy Technology Data Exchange (ETDEWEB)

Cliquet, Elisa; Ruault, Jean-Marc; Masson, Frederic, E-mail: elisa.cliquet@cnes.fr, E-mail: frederic.masson@cnes.fr [Centre National d' Etudes Spatiales (CNES), Paris (France); Roux, Jean-Pierre; Paris, Nicolas; Cazale, Brice; Manifacier, Laurent, E-mail: jean-pierre.roux@areva.com [AREVA TA, Aix en Provence, (France); Poinot-Salanon, Christine, E-mail: christine.poinot@cea.fr [Comissariado a l' Energie Atomique et Aux Energies alternatives (CEA), Paris (France)

2013-07-01

Nuclear propulsion has been studied for many decades. The power density of nuclear fission is much higher than chemical process, and for missions to outer solar system requiring several hundred of kilowatts, or for flexible manned missions to Mars requiring several megawatts, nuclear electric propulsion might be the only option offering a reasonable mass in low earth orbit. Despite the existence of low power experiences - SNAP10 in the 60's or Buk/Topaz in the 60-80's - no high power reactor has been developed: investment cost, long term time frame, high technological challenges and radioactive hazards are the main challenges we must overtake. However, it seems reasonable to look at the technical challenges that have to be overcome for a next generation of nuclear electric systems for space exploration. This paper will present some recent studies going on in France, on space reactors for exploration. Three classes of power have been considered: 10kWe, 100kWe, and several megawatts. Available data from previous studies and developments performed in Russia, USA], and Europe, have been collected and gave us a large overview of potential technical solutions. This was the starting point of a trade-off analysis aiming at the selection of the best options, with regards to the technological readiness level in France and Europe. The resulting preliminary designs will be presented and critical technologies needing maturation activities will be highlighted. (author)

Study of space reactors for exploration missions

International Nuclear Information System (INIS)

Cliquet, Elisa; Ruault, Jean-Marc; Masson, Frederic; Roux, Jean-Pierre; Paris, Nicolas; Cazale, Brice; Manifacier, Laurent; Poinot-Salanon, Christine

2013-01-01

Nuclear propulsion has been studied for many decades. The power density of nuclear fission is much higher than chemical process, and for missions to outer solar system requiring several hundred of kilowatts, or for flexible manned missions to Mars requiring several megawatts, nuclear electric propulsion might be the only option offering a reasonable mass in low earth orbit. Despite the existence of low power experiences - SNAP10 in the 60's or Buk/Topaz in the 60-80's - no high power reactor has been developed: investment cost, long term time frame, high technological challenges and radioactive hazards are the main challenges we must overtake. However, it seems reasonable to look at the technical challenges that have to be overcome for a next generation of nuclear electric systems for space exploration. This paper will present some recent studies going on in France, on space reactors for exploration. Three classes of power have been considered: 10kWe, 100kWe, and several megawatts. Available data from previous studies and developments performed in Russia, USA], and Europe, have been collected and gave us a large overview of potential technical solutions. This was the starting point of a trade-off analysis aiming at the selection of the best options, with regards to the technological readiness level in France and Europe. The resulting preliminary designs will be presented and critical technologies needing maturation activities will be highlighted. (author)

Space reactor safety, 1985--1995 lessons learned

International Nuclear Information System (INIS)

Marshall, A.C.

1995-01-01

Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration

Space reactor safety, 1985--1995 lessons learned

Energy Technology Data Exchange (ETDEWEB)

Marshall, A.C.

1995-12-31

Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration.

Closed Brayton cycle power conversion systems for nuclear reactors :

Energy Technology Data Exchange (ETDEWEB)

Wright, Steven A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lipinski, Ronald J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vernon, Milton E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Travis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2006-04-01

This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

Atomic Power in Space: A History

Science.gov (United States)

1987-03-01

"Atomic Power in Space," a history of the Space Isotope Power Program of the United States, covers the period from the program's inception in the mid-1950s through 1982. Written in non-technical language, the history is addressed to both the general public and those more specialized in nuclear and space technologies. Interplanetary space exploration successes and achievements have been made possible by this technology, for which there is no known substitue.

Multimegawatt space nuclear power open-cycle MHD-facility

International Nuclear Information System (INIS)

Pavshuk, V.A.; Panchenko, V.P.

2008-01-01

Paper presents the results of the efforts to calculate the characteristics, the layout and the engineering design of the open cycle space power propulsion on the basis of the high-temperature nuclear reactor for a nuclear rocket engine and the Faraday 20 MW capacity MHD-generator. The IVG-1 heterogeneous channel-vessel reactor ensuring in the course of the experiments hydrogen heating up to 3100 K, up to 5 MPa pressure at the reactor core outlet, up to 5 kg/s flowsheet, up to 220 MW thermal power served as a reactor is considered. One determined the MHD-generator basic parameters, namely: the portion of Cs dope was equal to 20%, the outlet stagnation pressure - 2 MPa, the electric conductivity - ≅30 S/m, the Mach number - ≅0.7, the magnetic field induction - 6 T, the capacity - 20 MW, the specific power removal - ∼4 MJ/kg. Paper describes the design of the MHD-facility with the working fluid momentless discharge and its basic characteristics [ru

Reactor power control method upon accidents of electrical power system

International Nuclear Information System (INIS)

Hirose, Masao.

1983-01-01

Purpose: To enable to continue the operation of a BWR type reactor by avoiding the scram while suppressing the reactor power, just after the external disturbance such as earth-trouble in power-transmission network. Method: Steep power drop of an electrical generator is to be detected not only by a current-type power-load-unbalance relay but also with a power-type power-load-unbalance-relay. If steep power-drop was detected by the latter relay, a previously selected control rod is rapidly inserted into the reactor. In this way, in the case where there is a possibility of the reactor scram, the scram can be avoided by suppressing the reactor power, thus the reactor operation can be continued. (Kamimura, M.)

Power reactors in member states

International Nuclear Information System (INIS)

1975-01-01

This is the first issue of a periodical computer-based listing of civilian nuclear power reactors in the Member States of the IAEA, presenting the situation as of 1 April 1975. It is intended as a replacement for the Agency's previous annual publication of ''Power and Research Reactors in Member States''. In the new format, the listing contains more information about power reactors in operation, under construction, planned and shut down. As far as possible all the basic design data relating to reactors in operation have been included. In future these data will be included also for other power reactors, so that the publication will serve to give a clear picture of the technical progress achieved. Test and research reactors and critical facilities are no longer listed. Of interest to nuclear power planners, nuclear system designers, nuclear plant operators and interested professional engineers and scientists

Method of estimating thermal power distribution of core of BWR type reactor

International Nuclear Information System (INIS)

Sekimizu, Koichi

1982-01-01

Purpose: To accurately and rapidly predict the thermal power of the core of a BWR they reactor at load follow-up operating time. Method: A parameter value corrected from a correction coefficient deciding unit and a xenon density distribution value predicted and calculated from a xenon density distributor are inputted to a thermal power distribution predicting devise, the status amount such as coolant flow rate or the like predetermined at this and next high power operating times is substituted for physical model to predict and calculate the thermal power distribution. The status amount of a nuclear reactor at the time of operating in previous high power corresponding to the next high power operation to be predicted is read from the status amount of the reactor stored in time series manner is a reactor core status memory, and the physical model used in the prediction and calculation of the thermal power distribution at the time of next high power operation is corrected. (Sikiya, K.)

About choosing the power unit of NPP

International Nuclear Information System (INIS)

Alekseev, P.N.; Mordashev, V.M.; Proshkin, A.A.; Stukalov, V.A.; Subbotin, S.A.; Tsibul'skij, V.F.; Chernilin, Yu.F.

2008-01-01

The present-day domestic nuclear power industry faces the necessity to make both process and economy related decisions. The mentioned decisions should cover all the aspects of the nuclear power complex. In particular, as of now, no substantiation of the present-day and the future requirements for the power unit of the WWER and fast reactor NPPs is available. The choice of the unit power of an NPP should involve consideration of all factors and should not be boiled down to the efforts to ensure the minimization of the investment and the operation constituents of the electric power generation costs. The aim of the present paper was to ensure elaboration of the recommendations to substantiate the optimal unit power of NPP power units based on the analysis of various factors to design, to construct and to operate an NPP (investment, time of construction, the unscheduled shutdown losses, the unification, the fabrication quality, the accident damage, etc.), and the electrical network potentialities and the regional electric power demands [ru

14 CFR 23.1142 - Auxiliary power unit controls.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Auxiliary power unit controls. 23.1142... Powerplant Controls and Accessories § 23.1142 Auxiliary power unit controls. Means must be provided on the... power unit. [Doc. No. 26344, 58 FR 18974, Apr. 9, 1993] ...

Reactor power reduction system and method

International Nuclear Information System (INIS)

Bruno, S.J.; Dunn, S.A.; Raber, M.

1978-01-01

A method of operating a nuclear power reactor is disclosed which enables an accelerated power reduction of the reactor without completely shutting the reactor down. The method includes monitoring the incidents which, upon their occurrence, would require an accelerated power reduction in order to maintain the reactor in a safe operation mode; calculating the power reduction required on the occurrence of such an incident; determining a control rod insertion sequence for the normal operation of the reactor, said sequence being chosen to optimize reactor power capability; selecting the number of control rods necessary to respond to the accelerated power reduction demand, said selection being made according to a priority determined by said control rod insertion sequence; and inserting said selected control rods into the reactor core. 11 claims, 13 figures

Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

International Nuclear Information System (INIS)

Boldyrev, V.M.; Mikhaj, V.I.

1985-01-01

Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

Survey of thorium utilization in power reactor systems

International Nuclear Information System (INIS)

Schwartz, M.H.; Schleifer, P.; Dahlberg, R.C.

1976-01-01

It is clear that thorium-fueled thermal power reactor systems based on current technology can play a vital role in serving present and long-term energy needs. Advanced thorium converters and thermal breeders can provide an expanded resource base from which the world's growing energy demands can be met. Utilization of a symbiotic system of fast breeders and thorium-fueled thermal reactors can be particularly effective in providing low cost power while conserving uranium resources. Breeder reactors are characterized by high capital costs and very low fuel costs since they produce more fuel than they consume. This excess fuel can be used to fuel thermal converter reactors whose capital costs are low. This symbiosis is optimized when 233 U is bred in the fast breeders and then used to fuel high-conversion-ratio thermal converter reactors operating on the thorium-uranium fuel cycle. The thorium-cycle HTGR, after undergoing more than fifteen years of development in both the United States and Europe, provides for the optimum utilization of our limited uranium resources. Other thermal reactor systems, previously operating on the uranium cycle, also show potential in their capability to utilize the thorium cycle effectively

Space reactor electric systems: system integration studies, Phase 1 report

International Nuclear Information System (INIS)

Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

1983-01-01

This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied

Feasible reactor power cutback logic development for an integral reactor

International Nuclear Information System (INIS)

Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

2013-01-01

Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

Reactor power distribution monitor

International Nuclear Information System (INIS)

Hoizumi, Atsushi.

1986-01-01

Purpose: To grasp the margin for the limit value of the power distribution peaking factor inside the reactor under operation by using the reactor power distribution monitor. Constitution: The monitor is composed of the 'constant' file, (to store in-reactor power distributions obtained from analysis), TIP and thermocouple, lateral output distribution calibrating apparatus, axial output distribution synthesizer and peaking factor synthesizer. The lateral output distribution calibrating apparatus is used to make calibration by comparing the power distribution obtained from the thermocouples to the power distribution obtained from the TIP, and then to provide the power distribution lateral peaking factors. The axial output distribution synthesizer provides the power distribution axial peaking factors in accordance with the signals from the out-pile neutron flux detector. These axial and lateral power peaking factors are synthesized with high precision in the three-dimensional format and can be monitored at any time. (Kamimura, M.)

Low-cost space fission power systems utilizing US and former Soviet Union experience and technology

International Nuclear Information System (INIS)

Wetch, J.R.; Britt, E.J.; Koester, J.K.; Gunther, N.; Ponomarev-Stepnoi, N.N.; Nikolaev, Y.V.; Nikitin, V.

1997-01-01

This report summarizes the author close-quote s approach to space power total economics. In the past 40 years of U.S. government sponsored space nuclear power developments, total economics has received only token consideration. In the real world, nuclear power has had limited acceptance where it provided the enabling capability i.e. isotopes for low power, long life, deep space missions, or reactor power for underwater nuclear submarines. It was also accepted where it was perceived to be more economic. Examples are nuclear reactor powered aircraft carriers, escort vessels and central station power stations. In any case, real and perceived public and environmental safety must always be included into the economic equation. copyright 1997 American Institute of Physics

Space power plants and power-consuming industrial systems

International Nuclear Information System (INIS)

Latyshev, L.; Semashko, N.

1996-01-01

An opportunity to create the space power production on the basis of solar, nuclear and fusion energies is analyzed. The priority of solar power production as the most accessible and feasible in comparison with others is emphasized. However, later on, it probably will play an auxiliary role. The possibilities of fusion power production, as a basic one in future, are also considered. It is necessary to create reactors using the fueling cycle with helium-3 (instead of tritium and deuterium, later on). The reaction products--charged particles, mainly--allow one to organize the system of direct fusion energy conversion into electricity. The produced energy is expected not to be transmitted to Earth, but an industry in space is expected to be produced on its basis. The industrial (power and science-consuming) objects located on a whole number of space apparatus will form a single complex with its own basic power plant. The power transmission within the complex will be realized with high power density fluxes of microwave radiation to short distances with their receivers at the objects. The necessary correction of the apparatus positions in the complex will be done with ion and plasma thrusters. The materials present on the Moon, asteroids and on other planets can serve as raw materials for industrial objects. Such an approach will help to improve the ecological state on Earth, to eliminate the necessity in the fast energy consumption growth and to reduce the hazard of global thermal crisis

12th Symposium on Space Nuclear Power and Propulsion. Conference on Alternative Power from Space (APFS),Conference on Accelerator-Driven Transmutation Technologies and Applications (A-DTTA)

International Nuclear Information System (INIS)

Mohamed, S.E.

1995-01-01

These proceedings represent papers presented at the 12th symposium on Space Nuclear Power and Propulsion held in Albuquerque, New Mexico. The symposium theme was ''commercialization and technology transfer''. The topics discussed include: wireless power transmission, solar power from space next generation spacecraft, space power electronics and power management, flight testing of components, manufacturing and processing of materials, nuclear propulsion, reactors and shielding and many others of interest to the scientific community representing industry, government and academic institutions. There were 163 papers presented at the conference and 60 have been abstracted for the Energy Science and Technology database

Materialistic Aspects of Raising Resource of Pressurized Water Reactors for Low-Power Nuclear Plants

International Nuclear Information System (INIS)

Parshin, A.M.; Muratov, O.E.

2005-01-01

The opportunity of using ships reactors for low-power nuclear plants is considered. Some aspects of working constructional materials on cases of water-water reactors of ships nuclear units are considered. Advantages of raising resource of ships reactors are shown

Renovation of the 'old' NPP units as an economically effective way of nuclear power development

International Nuclear Information System (INIS)

Zrodnikov, A.V.; Toshinsky, G.I.; Komlev, O.G.; Dragunov, Yu.G.; Stepanov, V.S.; Klimov, N.N.; Kopytov, I.I.; Krushelnitsky, V.N.

2005-01-01

In the process of nuclear power development there comes a phase when the old' power-units, which reactor installations have expired the designed and extended service lifetime are withdrawn from operating. At this phase in the case of the same annual investments into nuclear power, the increase of the total set up capacity of the nuclear power will be terminated because introduction of capacities due to construction of 'new' power-units only compensates for the reduction of capacities caused by withdrawing from operating the 'old' power-units. Along with this, taking into account a sizeable difference in the service lifetime of the nuclear steam supplying systems and the rest infrastructure of the nuclear power plants, it is an opportunity to find the solution to the problem of compensating for the withdrawn capacities without considerable increasing the annual investments. This opportunity is connected with use of the innovative nuclear power technology based on multipurpose small power modular fast reactors with lead-bismuth coolant for replacement of the withdrawn capacities (renovation of power-units). The features of the innovative nuclear power technology based on the SVBR-76/100 reactor installations, the results of the technical and economical investigations that demonstrate the high economical efficiency of use of the renovation technology using the SVBR-75/100 reactor modules are presented in the Paper. (author)

A new concept for filtration units for trapping radioactive aerosols and iodine in the ventilation systems of nuclear power plants with WWER reactors

International Nuclear Information System (INIS)

Foerster, V.; Slanina, S.

1985-01-01

The paper describes a concept for new filtration units in the ventilation systems of nuclear power plants with WWER reactors. The new units are characterized by more stringent requirements on the efficiency of air purification (removal of radioactive contaminants) and various requirements for the quality of air purification in the ventilation systems. Work performed at the Scientific Research Institute for Air Technology has resulted in filtration units of a universal modular type, the structural design of which permits a high degree of variation in their component parts. A brief description is given of the filtration units, their basic technical characteristics and examples of their use in nuclear power plant ventilation systems. (author)

Underground nuclear power station using self-regulating heat-pipe controlled reactors

International Nuclear Information System (INIS)

Hampel, V.E.

1989-01-01

The author presents a nuclear reactor for generating electricity disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor

Underground nuclear power station using self-regulating heat-pipe controlled reactors

Science.gov (United States)

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Status of NASA's Stirling Space Power Converter Program

International Nuclear Information System (INIS)

Dudenhoefer, J.E.; Winter, J.M.

1994-01-01

An overview is presented of the NASA Lewis Research Center Free-Piston Stirling Space Power Converter Technology Program. This work is being conducted under NASA's Civil Space Technology Initiative. The goal of the CSTI High Capacity Power Element is to develop the technology base needed to meet the long duration, high capacity power requirements for future NASA space initiatives. Efforts are focused upon increasing system power output and system thermal and electric energy conversion efficiency at least fivefold over current SP-100 technology, and on achieving systems that are compatible with space nuclear reactors. This paper will discuss Stirling experience in Space Power Converters. Fabrication is nearly completed for the 1050 K Component Test Power Converter (CTPC); results of motoring tests of the cold end (525 K), are presented. The success of these and future designs is dependent upon supporting research and technology efforts including heat pipes, bearings, superalloy joining technologies, high efficiency alternators, life and reliability testing and predictive methodologies. This paper provides an update of progress in some of these technologies leading off with a discussion of free-piston Stirling experience in space

Reactor power control system

International Nuclear Information System (INIS)

Tomisawa, Teruaki.

1981-01-01

Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

Flux distribution measurements in the Bruce A unit 1 reactor

International Nuclear Information System (INIS)

Okazaki, A.; Kettner, D.A.; Mohindra, V.K.

1977-07-01

Flux distribution measurements were made by copper wire activation during low power commissioning of the unit 1 reactor of the Bruce A generating station. The distribution was measured along one diameter near the axial and horizontal midplanes of the reactor core. The activity distribution along the copper wire was measured by wire scanners with NaI detectors. The experiments were made for five configurations of reactivity control mechanisms. (author)

NUCLEAR THERMIONIC SPACE POWER SYSTEMS

Energy Technology Data Exchange (ETDEWEB)

Howard, R. C.; Rasor, N. S.

1963-03-15

The various concepts for utilizing thermionic conversion in space reactor power plants are described and evaluated. The problems (and progress toward their solution) of the in-core concept, particularly, are considered. Progress in thermionic conversion technology is then reviewed from both the hardware and research points of view. Anticipated progress in thermionic conversion and the possible consequences for the performance of electrical propulsion systems are summarized. 46 references. (D.C.W.)

Handling of views and opinions by staters and others in a public hearing on alteration in reactor installation (addition of Unit 2) in the Sendai Nuclear Power Station of Kyushu Electric Power Co., Inc

International Nuclear Information System (INIS)

1981-01-01

A public hearing on the addition of Unit 2 in the Sendai Nuclear Power Station, Kyushu Electric Power Co., Inc., was held on July 17, 1980, in Sendai City, Kagoshima Prefecture. The views and opinions by the local staters and those by the notification of statement were expressed concerning its nuclear safety. The handling of these views and opinions by the Nuclear Safety Commission is explained. The most important in this action is the instruction by the NSC to the Committee on Examination of Reactor Safety to reflect the results of the public hearing to the reactor safety examination of the Unit 2 installation by the CERS. The views and opinions expressed in this connection are summarized as follows: the sitting conditions, the safety design of the reactor plant, and the release of radioactive materials, involving such aspects as earthquakes, accidents and radioactive waste management. (J.P.N.)

Effects of irradiation on properties of refractory alloys with emphasis on space power reactor applications

International Nuclear Information System (INIS)

Wiffen, F.W.

1984-01-01

The probable effects of irradiation on niobium and tungsten alloys in use as components of thermionic convertors in a space reactor were reviewed by the author in 1971. While considerably more data on refractory metals have been generated since that time, the data have not been reviewed with respect to space reactor applications. This paper attempts such a review. The approach used is to work from the most recently available review of irradiation effects for each alloy system (where such a review is available) and to discuss that review and more recent data judged to be the most useful in establishing likely behavior in high-temperature reactor service. 28 figures, 6 tables

Low power modular power generating reactors or Small Modular Reactors (SMR)

International Nuclear Information System (INIS)

Chenais, Jacques

2016-01-01

Electronuclear reactors were small reactors at the beginning, and then tend to be always bigger and more powerful, but since some recent times, several countries specialized in reactor design and fabrication (USA, Russia, China, and South Korea) have been developing Small Modular Reactors (SMR) of less than 300 MW. As France has already produced feasibility studies and is about to launch a SMR development programme, the author comments some specific aspects of this new architecture of reactors, characterises the targeted markets, gives an overview of the various more or less advanced existing concepts: a floating barge in Russia, the SMART 100 MW project in South Korea, several concepts in the USA (the mPower 125 MW, the NuScale 45 MW, the Westinghouse 225 MW, and the HI-SMUR 160 MW projects), the ACP 100 MW in China, the CAREM 27 MW in Argentina. French projects developed by the CEA, EDF, Areva and DCNS are then presented

Identification of fast power reactivity effect in nuclear power reactor

International Nuclear Information System (INIS)

Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

1987-01-01

A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2014-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

A nuclear power reactor

International Nuclear Information System (INIS)

Borrman, B.E.; Broden, P.; Lundin, N.

1979-12-01

The invention consists of shock absorbing support beams fastened to the underside of the reactor tank lid of a BWR type reactor, whose purpose is to provide support to the steam separator and dryer unit against accelerations due to earthquakes, without causing undue thermal stresses in the unit due to differential expansion. (J.I.W.)

Power oscillations in BWR reactors

International Nuclear Information System (INIS)

Espinosa P, G.

2002-01-01

One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

Aerosol core nuclear reactor for space-based high energy/power nuclear-pumped lasers

International Nuclear Information System (INIS)

Prelas, M.A.; Boody, F.P.; Zediker, M.S.

1987-01-01

An aerosol core reactor concept can overcome the efficiency and/or chemical activity problems of other fuel-reactant interface concepts. In the design of a laser using the nuclear energy for a photon-intermediate pumping scheme, several features of the aerosol core reactor concept are attractive. First, the photon-intermediate pumping concept coupled with photon concentration methods and the aerosol fuel can provide the high power densities required to drive high energy/power lasers efficiently (about 25 to 100 kW/cu cm). Secondly, the intermediate photons should have relatively large mean free paths in the aerosol fuel which will allow the concept to scale more favorably. Finally, the aerosol core reactor concept can use materials which should allow the system to operate at high temperatures. An excimer laser pumped by the photons created in the fluorescer driven by a self-critical aerosol core reactor would have reasonable dimensions (finite cylinder of height 245 cm and radius of 245 cm), reasonable laser energy (1 MJ in approximately a 1 millisecond pulse), and reasonable mass (21 kg uranium, 8280 kg moderator, 460 kg fluorescer, 450 kg laser medium, and 3233 kg reflector). 12 references

Power generator in BWR type reactors

International Nuclear Information System (INIS)

Yoshida, Kenji.

1984-01-01

Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

Analysis of the accident in the second power-generating unit of the Chernobyl nuclear power plant caused by inadequate makeup of the reactor cooling loop

International Nuclear Information System (INIS)

Vasil'chenko, V.N.; Kramerov, A.Ya.; Mikhailov, D.A.

1995-01-01

The accident in the second power-generating unit of the Chernobyl nuclear power plant on October 11, 1991 was the result of unauthorized connection of the TG-4 turbogenerator, which was shut down for repairs, into the grid (in the off-design asynchronous engine mode), and this resulted in a serious fire in the machine room and subsequent failure of systems which are important for safety and which ensure the design mode of reactor cooling: These were primarily failures of the feed and emergency feed pumps and failure of the BRU-B control valve, which regulates steam release during cooling

Space-time reactor kinetics for heterogeneous reactor structure

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1969-11-15

An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Design and Test Plans for a Non-Nuclear Fission Power System Technology Demonstration Unit

Science.gov (United States)

Mason, Lee; Palac, Donald; Gibson, Marc; Houts, Michael; Warren, John; Werner, James; Poston, David; Qualls, Arthur Lou; Radel, Ross; Harlow, Scott

2012-01-01

A joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) team is developing concepts and technologies for affordable nuclear Fission Power Systems (FPSs) to support future exploration missions. A key deliverable is the Technology Demonstration Unit (TDU). The TDU will assemble the major elements of a notional FPS with a non-nuclear reactor simulator (Rx Sim) and demonstrate system-level performance in thermal vacuum. The Rx Sim includes an electrical resistance heat source and a liquid metal heat transport loop that simulates the reactor thermal interface and expected dynamic response. A power conversion unit (PCU) generates electric power utilizing the liquid metal heat source and rejects waste heat to a heat rejection system (HRS). The HRS includes a pumped water heat removal loop coupled to radiator panels suspended in the thermal-vacuum facility. The basic test plan is to subject the system to realistic operating conditions and gather data to evaluate performance sensitivity, control stability, and response characteristics. Upon completion of the testing, the technology is expected to satisfy the requirements for Technology Readiness Level 6 (System Demonstration in an Operational and Relevant Environment) based on the use of high-fidelity hardware and prototypic software tested under realistic conditions and correlated with analytical predictions.

Applicability of trends in nuclear safety analysis to space nuclear power systems

International Nuclear Information System (INIS)

Bari, R.A.

1992-01-01

A survey is presented of some current trends in nuclear safety analysis that may be relevant to space nuclear power systems. This includes: lessons learned from operating power reactor safety and licensing; approaches to the safety design of advanced and novel reactors and facilities; the roles of risk assessment, extremely unlikely accidents, safety goals/targets; and risk-benefit analysis and communication

Fast reactors in nuclear power

Energy Technology Data Exchange (ETDEWEB)

Kazachkovskii, O

1981-02-01

The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

Free-piston Stirling technology for space power

International Nuclear Information System (INIS)

Slaby, J.G.

1994-01-01

An overview is presented of the NASA Lewis Research Center free-piston Stirling engine activities directed toward space power. This work is being carried out under NASA's new Civil Space Technology Initiative (CSTI). The overall goal of CSTI's High Capacity Power element is to develop the technology base needed to meet the long duration, high capacity power requirements for future NASA space missions. The Stirling cycle offers an attractive power conversion concept for space power needs. Discussed in this paper is the completion of the Space Power Demonstrator Engine (SPDE) testing - culminating in the generation of 25 kW of engine power from a dynamically-balanced opposed-piston Stirling engine at a temperature ratio of 2.0. Engine efficiency was approximately 22 percent. The SPDE recently has been divided into two separate single-cylinder engines, called Space Power Research Engines (SPRE), that now serve as test beds for the evaluation of key technology disciplines. These disciplines include hydrodynamic gas bearings, high-efficiency linear alternators, space qualified heat pipe heat exchangers, oscillating flow code validation, and engine loss understanding. The success of the SPDE at 650 K has resulted in a more ambitious Stirling endeavor - the design, fabrication, test and evaluation of a designed-for-space 25 kW per cylinder Stirling Space Engine (SSE). The SSE will operate at a hot metal temperature of 1050 K using superalloy materials. This design is a low temperature confirmation of the 1300 K design. It is the 1300 K free-piston Stirling power conversion system that is the ultimate goal; to be used in conjunction with the SP-100 reactor. The approach to this goal is in three temperature steps. However, this paper concentrates on the first two phases of this program - the 650 K SPDE and the 1050 K SSE

The Optimization of power reactor control system

International Nuclear Information System (INIS)

Danupoyo, S.D.

1997-01-01

A power reactor is an important part in nuclear powered electrical plant systems. Success in controlling the power reactor will establish safety of the whole power plant systems. Until now, the power reactor has been controlled by a classical control system that was designed based on output feedback method. To meet the safety requirements that are now more restricted, the recently used power reactor control system should be modified. this paper describes a power reactor control system that is designed based on a state feedback method optimized with LQG (Linear-quadrature-gaussian) method and equipped with a state estimator. A pressurized-water type reactor has been used as the model. by using a point kinetics method with one group delayed neutrons. the result of simulation testing shows that the optimized control system can control the power reactor more effective and efficient than the classical control system

Technical Bases to Consider for Performance and Demonstration Testing of Space Fission Reactors

International Nuclear Information System (INIS)

Hixson, Laurie L.; Houts, Michael G.; Clement, Steven D.

2004-01-01

Performance and demonstration testing are critical to the success of a space fission reactor program. However, the type and extent to which testing of space reactors should be performed has been a point of discussion within the industry for many years. With regard to full power ground nuclear tests, questions such as 'Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Will the test article accurately represent the flight system? Are the costs too restrictive?' have been debated for decades. There are obvious benefits of full power ground nuclear testing such as obtaining systems integrated reliability data on a full-scale, complete end-to-end system. But these benefits come at some programmatic risk. In addition, this type of testing does not address safety related issues. This paper will discuss and assess these and other technical considerations essential in deciding which type of performance and demonstration testing to conduct on space fission reactor systems. (authors)

Utilization of nuclear energy for generating electric power in the FRG, with special regard to LWR-type reactors

International Nuclear Information System (INIS)

Vollradt, J.

1977-01-01

Comments on interdependencies in energy industry and energy generation as seen by energy supply utilities, stating that the generation of electric power in Germany can only be based on coal and nuclear energy in the long run, are followed by the most important, fundamental, nuclear-physical, technological and in part political interdependencies prevailing in the starting situation of 1955/58 when the construction of nuclear power plant reactors began. Then the development ranging to the 28000 MW nuclear power output to be expected in 1985 is outlined, totalling in 115000 MW electric power in the FRG. Finally, using the respectively latest order, the technical set up of each of the reactor types with 1300 MWe unit power offered by German manufacturers are described: BBC/BBR PWR-type reactor Neupotz, KWU-PWR-type reactor Hamm and KWU PWR-type reactor double unit B+C Gundremmingen. (orig.) [de

Transactions of the fifth symposium on space nuclear power systems

Energy Technology Data Exchange (ETDEWEB)

El-Genk, M.S.; Hoover, M.D. (eds.)

1988-01-01

This paper contains the presented papers at the fourth symposium on space nuclear power systems. Topics of these paper include: space nuclear missions and applications, reactors and shielding, nuclear electric and nuclear propulsion, high-temperature materials, instrumentation and control, energy conversion and storage, space nuclear fuels, thermal management, nuclear safety, simulation and modeling, and multimegawatt system concepts. (LSP)

Space Nuclear Reactor Engineering

Energy Technology Data Exchange (ETDEWEB)

Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-03-06

We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

TerraPower, Bill Gates' reactor

International Nuclear Information System (INIS)

Guidez, J.

2016-01-01

TerraPower is a traveling wave reactor, it means that the reactor gradually converts non fissile material into the fuel it needs and the active part of the core progressively moves through the core leaving spent fuel behind. The last design of the TerraPower shows that it will use depleted uranium as fuel and that its core will need reloading every 10 years. Re-arrangement of the nuclear fuel will have to be made every 18 months to keep the core reactive. Metallic nuclear fuels will be used as they allow the highest breeding rates. It appears that apart from the very specific configuration of the core, the TerraPower is a reactor very similar to sodium-cooled fast reactors. Neutron transport inside traveling wave reactor core is complex and simulations show that the piling-up of fission product tends to kill the chain reaction and a continuous neutron addition may be necessary to keep the reactor going. A large part of the TerraPower feasibility studies concerns neutron transport inside its core. (A.C.)

Power reactor noise

International Nuclear Information System (INIS)

Thie, J.A.

1981-01-01

This book concentrates on the different types of noise present in power reactors and how the analysis of this noise can be used as a tool for reactor monitoring and diagnostics. Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions thus preventing further complications by alerting operators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussion of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

Development of a power-period calculation unit for nuclear reactor Control

International Nuclear Information System (INIS)

Martin, J.

1966-10-01

The apparatus studied is a digital calculating assembly which makes it possible to prepare and to present numerically the period and power of a nuclear reactor during operation, from start-up to nominal power. The pulses from a fission chamber are analyzed continuously, using real time. A small number of elements is required because of the systematic use of a calculation technique comprising the determination of a base 2 logarithm by a linear approximation. The accuracy obtained for the period is of the order of 14%; the response time of the order of the calculated period value. An approximate value of the power (30%) is given at each calculation cycle together with the power thresholds required for the control. (author) [fr

Lightweight power bus for a baseload nuclear reactor in space

International Nuclear Information System (INIS)

Oberly, C.E.; Massie, L.D.; Hoffman, D.J.

1989-01-01

Space environmental interactions with the power distribution/power processing subsystem can become a serious problem for power systems rated at 10's to 100's of kilowatts. Utilization of ceramic superconductors at 1000 A/cm/sup 2/, which has already been demonstrated at 77 K in a conductor configuration may eliminate both bus mass and distribution voltage problems in a high power satellite. The analytical results presented here demonstrate that a superconducting coaxial power transmission bus offers significant benefits in reduced distribution voltage and mass

Nuclear Space Power Systems Materials Requirements

International Nuclear Information System (INIS)

Buckman, R.W. Jr.

2004-01-01

High specific energy is required for space nuclear power systems. This generally means high operating temperatures and the only alloy class of materials available for construction of such systems are the refractory metals niobium, tantalum, molybdenum and tungsten. The refractory metals in the past have been the construction materials selected for nuclear space power systems. The objective of this paper will be to review the past history and requirements for space nuclear power systems from the early 1960's through the SP-100 program. Also presented will be the past and present status of refractory metal alloy technology and what will be needed to support the next advanced nuclear space power system. The next generation of advanced nuclear space power systems can benefit from the review of this past experience. Because of a decline in the refractory metal industry in the United States, ready availability of specific refractory metal alloys is limited

Correlations between power and test reactor data bases

International Nuclear Information System (INIS)

Guthrie, G.L.; Simonen, E.P.

1989-02-01

Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

Reactor power distribution monitor

International Nuclear Information System (INIS)

Sekimizu, Koichi

1980-01-01

Purpose: To improve the performance and secure the safety of a nuclear reactor by rapidly computing and display the power density in the nuclear reactor by using a plurality of processors. Constitution: Plant data for a nuclear reactor containing the measured values from a local power monitor LPRM are sent and recorded in a magnetic disc. They are also sent to a core performance computer in which burn-up degree distribution and the like are computed, and the results are sent and recorded in the magnetic disc. A central processors loads programs to each of the processors and applies data recorded in the magnetic disc to each of the processors. Each of the processors computes the corresponding power distribution in four fuel assemblies surrounding the LPRM string by the above information. The central processor compiles the computation results and displays them on a display. In this way, power distribution in the fuel assemblies can rapidly be computed to thereby secure the improvement of the performance and safety of the reactor. (Seki, T.)

Nuclear power supplies: their potential and the practical problems to their achievement for space missions

International Nuclear Information System (INIS)

Colston, B.W.; Brehm, R.L.

1985-01-01

The anticipated growth of the space station power requirement provides a good example of the problem the space nuclear power supply developers have to contend with: should a reactor power supply be developed that attempts to be all things to all missions, i.e., is highly flexible in its ability to meet a wide variety of missions, or should the development of a reactor system await a specific mission definition and be customized to this mission. This leads, of course, to a chicken-and-egg situation. For power requirements of several hundreds of kilowatts or more, no nuclear power source exists or is even far enough along in the definition stage (much less the development stage) for NASA to reasonably assume probable availability within the next 10 years. The real problem of space nuclear power is this ''chicken-and-egg'' syndrome: DOE will not develop a space reactor system for NASA without a firm mission, and NASA will not specify a firm mission requiring a space reactor because such a system doesn't exist and is perceived not to be developable within the time frame of the mission. The problem is how to break this cycle. The SP-100 program has taken an important first step to breaking this cycle, but this program is much more design-specific than what is required to achieve a broad technology base and latitude in achievable power level. In contrast to the SP-100 approach, a wider perspective is required: the development of the appropriate technologies for power levels can be broken into ranges, say, from 100 kWe to 1000 kWe, and from 1000 kWe to 10,000 kWe

Safety features of a power unit with the BREST-OD-300 reactor

International Nuclear Information System (INIS)

Smirnov, V.S.

2013-01-01

• Two classes of the most severe accidents have been considered: with positive reactivity introduction and with loss of forced heat removal from the core. • It has been demonstrated that no full reactivity margin introduction during reactor power operations leads to the temperature limits of the fuel element safe operation being exceeded when the EP system or only the Passive EP-T system operates. • It has been shown that no loss of forced heat removal during blackout results in the temperature limits of safe operation being exceeded when at least only passive protections (Passive EP-T and PFBS) operate. • It has been shown that no even most severe accidents involving full reactivity margin introduction (UTOP) and full loss of power for all reactor facility systems (ULOF) lead to fuel melting and coolant boiling

An underground nuclear power station using self-regulating heat-pipe controlled reactors

Science.gov (United States)

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Maximizing the use of research reactors in training power reactor operating staff with special reference to US experience

International Nuclear Information System (INIS)

Cox, J.A.

1976-01-01

Research reactors have been used in training nuclear power plant personnel for many years. Using the experience in the United States of America a programme is proposed that will maximize the training conducted at a research reactor and lessen the time that the staff must spend training elsewhere. The programme is adaptable to future training of replacement staff and for staff retraining. (author)

Annex VII - Diagrams: 1. Reactor operation (1960-1977); 2. Mean daily reactor power density in 1977; 3. Monthly reactor power for 1977; 4. percent of utilization of experimental space in 1977; Prilog VII - Dijagrami: 1. Rad reaktora (MWh) po godinama (1960-1977); 2. Srednja dnevna snaga reaktora u 1977. godini; 3. Rad reaktora (MWh) po mesecima za 1977. godinu i 4. Procenat iskoriscenja eksperimentalnog prostora u 1977. godini

Energy Technology Data Exchange (ETDEWEB)

NONE

1977-12-15

This Annex includes the following diagrams: 1. Annual Reactor RA power production (MWh) for the period from 1960-1977; 2. Mean daily reactor power density MW in 1977; 3. Monthly reactor power production (MWh) for 1977; 4. percent of utilization of experimental space in 1977. [Serbo-Croat] Ovaj prilog sadrzi dijagrame: 1. Rad reaktora (MWh) po godinama (1960-1977); 2. Srednja dnevna snaga reaktora u 1977. godini; 3. Rad reaktora (MWh) po mesecima za 1977. godinu i 4. Procenat iskoriscenja eksperimentalnog prostora u 1977. godini.

Nuclear power station with nuclear reactor accommodated largely secure against catastrophes

International Nuclear Information System (INIS)

Rosen, O.

1987-01-01

If the nuclear reactor is installed underground near the power station unit, then danger to the environment due to radiation contamination can be largely or nearly completely prevented by a covering of constant thickness or by a covering which can be installed by a catastrophic accident. The extinguishing of a burning reactor is also relatively simple for a reactor accommodated in a pit. The above-mentioned measures can be used individually or combined. (orig./HP) [de

The effects of electric power industry restructuring on the safety of nuclear power plants in the United States

Science.gov (United States)

Butler, Thomas S.

Throughout the United States the electric utility industry is restructuring in response to federal legislation mandating deregulation. The electric utility industry has embarked upon an extraordinary experiment by restructuring in response to deregulation that has been advocated on the premise of improving economic efficiency by encouraging competition in as many sectors of the industry as possible. However, unlike the telephone, trucking, and airline industries, the potential effects of electric deregulation reach far beyond simple energy economics. This dissertation presents the potential safety risks involved with the deregulation of the electric power industry in the United States and abroad. The pressures of a competitive environment on utilities with nuclear power plants in their portfolio to lower operation and maintenance costs could squeeze them to resort to some risky cost-cutting measures. These include deferring maintenance, reducing training, downsizing staff, excessive reductions in refueling down time, and increasing the use of on-line maintenance. The results of this study indicate statistically significant differences at the .01 level between the safety of pressurized water reactor nuclear power plants and boiling water reactor nuclear power plants. Boiling water reactors exhibited significantly more problems than did pressurized water reactors.

Primary shutdown system monitoring unit for nuclear power plants

International Nuclear Information System (INIS)

Khan, Tahir Kamal; Balasubramanian, R.; Agilandaeswari, K.

2013-01-01

Shut off rods made up of neutron absorbing material are used as Primary Shutdown System. To reduce the power of the reactor under certain abnormal operating conditions, these rods must go down into the core within a specified time. Any malfunctioning in the movement of rods cannot be tolerated and Secondary Shutdown System (SSS) must be actuated within stipulated time to reduce the reactor power. A special safety critical, hardwired electronics unit has been designed to detect failure of PSS Shut off rods movements and generate trip signals for initiating SSS. (author)

Concepts for space nuclear multi-mode reactors

International Nuclear Information System (INIS)

Myrabo, L.; Botts, T.E.; Powell, J.R.

1983-01-01

A number of nuclear multi-mode reactor power plants are conceptualized for use with solid core, fixed particle bed and rotating particle bed reactors. Multi-mode systems generate high peak electrical power in the open cycle mode, with MHD generator or turbogenerator converters and cryogenically stored coolants. Low level stationkeeping power and auxiliary reactor cooling (i.e., for the removal of reactor afterheat) are provided in a closed cycle mode. Depending on reactor design, heat transfer to the low power converters can be accomplished by heat pipes, liquid metal coolants or high pressure gas coolants. Candidate low power conversion cycles include Brayton turbogenerator, Rankine turbogenerator, thermoelectric and thermionic approaches. A methodology is suggested for estimating the system mass of multi-mode nuclear power plants as a function of peak electric power level and required mission run time. The masses of closed cycle nuclear and open cycle chemical power systems are briefly examined to identify the regime of superiority for nuclear multi-mode systems. Key research and technology issues for such power plants are also identified

Proposed advanced satellite applications utilizing space nuclear power systems

International Nuclear Information System (INIS)

Bailey, P.G.; Isenberg, L.

1990-01-01

A review of the status of space nuclear reactor systems and their possible applications is presented. Such systems have been developed over the past twenty years and are capable of use in various military and civilian applications in the 5-1000 kWe power range. The capabilities and limitations of the currently proposed nuclear reactor systems are summarized. Safety issues are shown to be identified, and if properly addressed should not pose a hindrance. Applications are summarized for the federal and civilian community. These applications include both low and high altitude satellite surveillance missions, communications satellites, planetary probes, low and high power lunar and planetary base power systems, broad-band global telecommunications, air traffic control, and high-definition television

Transactions of the fourth symposium on space nuclear power systems

Energy Technology Data Exchange (ETDEWEB)

El-Genk, M.S.; Hoover, M.D. (eds.)

1987-01-01

This paper contains the presented papers at the fourth symposium on space nuclear power systems. Topics of these papers include: space nuclear missions and applications, reactors and shielding, nuclear electric and nuclear propulsion, refractory alloys and high-temperature materials, instrumentation and control, energy conversion and storage, space nuclear fuels, thermal management, nuclear safety, simulation and modeling, and multimegawatt system concepts. (LSP)

Mars, the Moon, and the Ends of the Earth: Autonomy for Small Reactor Power Systems

International Nuclear Information System (INIS)

Wood, Richard Thomas

2008-01-01

In recent years, the National Aeronautics and Space Administration (NASA) has been considering deep space missions that utilize a small-reactor power system (SRPS) to provide energy for propulsion and spacecraft power. Additionally, application of SRPS modules as a planetary power source is being investigated to enable a continuous human presence for nonpolar lunar sites and on Mars. A SRPS can supply high-sustained power for space and surface applications that is both reliable and mass efficient. The use of small nuclear reactors for deep space or planetary missions presents some unique challenges regarding the operations and control of the power system. Current-generation terrestrial nuclear reactors employ varying degrees of human control and decision-making for operations and benefit from periodic human interaction for maintenance. In contrast, the control system of a SRPS employed for deep space missions must be able to accommodate unattended operations due to communications delays and periods of planetary occlusion while adapting to evolving or degraded conditions with no opportunity for repair or refurbishment. While surface power systems for planetary outposts face less extreme delays and periods of isolation and may benefit from limited maintenance capabilities, considerations such as human safety, resource limitations and usage priorities, and economics favor minimizing direct, continuous human interaction with the SRPS for online, dedicated power system management. Thus, a SRPS control system for space or planetary missions must provide capabilities for operational autonomy. For terrestrial reactors, large-scale power plants remain the preferred near-term option for nuclear power generation. However, the desire to reduce reliance on carbon-emitting power sources in developing countries may lead to increased consideration of SRPS modules for local power generation in remote regions that are characterized by emerging, less established infrastructures

Preoperation of Hamaoka Nuclear Power Station Unit No. 4

International Nuclear Information System (INIS)

Fukuyo, Tadashi; Kurata, Satoshi

1994-01-01

Chubu Electric Power Co. finished preoperation of Hamaoka Nuclear Power Station Unit No. 4 in September, 1993. Although unit 4 has the same reactor design as unit 3, its rated electrical output (1,137MW) is 37MW more than that of unit 3. This increase was achieved mainly by adopting a Moisture Separater Heater in the turbine system. We started preoperation of unit 4 in November 1992 and performed various tests at electrical outputs of 20%, 50%, 75%, and 100%. We finished preoperation without any scram or other major problems and obtained satisfactory results for the functions and performance of the plant. This paper describes the major results of unit 4 preoperation. (author)

Heat pipe nuclear reactor for space power

Science.gov (United States)

Koening, D. R.

1976-01-01

A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

Statistical analysis of the vibration loading of the reactor internals and fuel assemblies of reactor units type WWER-440 from deferent projects

International Nuclear Information System (INIS)

Ovcharov, O.; Pavelko, V.; Usanov, A.; Arkadov, G.; Dolgov, A.; Molchanov, V.; Anikeev, J.; Pljush, A.

2006-01-01

In this paper the following items have been presented: 1) Vibration noise instrument channels; 2) Vibration loading characteristics of control assemblies, internals and design peculiarities of internals of WWER-440 deferent projects; 3) Coolant flow rate through the reactor, reactor core, fuel assemblies and control assemblies for different projects WWER-440 and 4) Noise measurements of coolant speed per channel. The change of auto power spectrum density of absolute displacement detector signal for the last 12 years of SUS monitoring of the Kola NPP unit 2; the coherence functions groups between two SPND of the same level for the Kola NPP unit 1; the measured coolant flow rate at Paks NPP and the auto power spectrum density group of SPND signals from 11 neutron measuring channels of the Kola NPP unit 1 are given. The main factors of vibration loading of internals and fuel assemblies for Kola NPP units 1-4, Bohunice NPP units 1 and 2 and Novovoronezh NPP units 3 and 4 are also discussed

Requirements under decree 430 UJD for unit outlet and standby power supply

International Nuclear Information System (INIS)

Vanco, K.

2012-01-01

At present, the nuclear power plants are only sufficient resources, which can cover a huge demand for electricity. Concentration so huge power in one place require adequate security from the perspective lead power outlet and standby power supply of reactor unit. (Author)

Applications of power beaming from space-based nuclear power stations. [Laser beaming to airplanes; microwave beaming to ground

Energy Technology Data Exchange (ETDEWEB)

Powell, J.R.; Botts, T.E.; Hertzberg, A.

1981-01-01

Power beaming from space-based reactor systems is examined using an advanced compact, lightweight Rotating Bed Reactor (RBR). Closed Brayton power conversion efficiencies in the range of 30 to 40% can be achieved with turbines, with reactor exit temperatures on the order of 2000/sup 0/K and a liquid drop radiator to reject heat at temperatures of approx. 500/sup 0/K. Higher RBR coolant temperatures (up to approx. 3000/sup 0/K) are possible, but gains in power conversion efficiency are minimal, due to lower expander efficiency (e.g., a MHD generator). Two power beaming applications are examined - laser beaming to airplanes and microwave beaming to fixed ground receivers. Use of the RBR greatly reduces system weight and cost, as compared to solar power sources. Payback times are a few years at present prices for power and airplane fuel.

Flux distribution measurements in the Bruce B Unit 6 reactor using a transportable traveling flux detector system

International Nuclear Information System (INIS)

Leung, T.C.; Drewell, N.H.; Hall, D.S.; Lopez, A.M.

1987-01-01

A transportable traveling flux detector (TFD) system for use in power reactors has been developed and tested at Chalk River Nuclear Labs. in Canada. It consists of a miniature fission chamber, a motor drive mechanism, a computerized control unit, and a data acquisition subsystem. The TFD system was initially designed for the in situ calibration of fixed self-powered detectors in operating power reactors and for flux measurements to verify reactor physics calculations. However, this system can also be used as a general diagnostic tool for the investigation of apparent detector failures and flux anomalies and to determine the movement of reactor internal components. This paper describes the first successful use of the computerized TFD system in an operating Canada deuterium uranium (CANDU) power reactor and the results obtained from the flux distribution measurements. An attempt is made to correlate minima in the flux profile with the locations of fuel channels so that future measurements can be used to determine the sag of the channels. Twenty-seven in-core flux detector assemblies in the 855-MW (electric) Unit 6 reactor of the Ontario Hydro Bruce B Generating Station were scanned

Use mobile pumps and liquid chilling water units to provide chilled water for nuclear reactor during nuclear power plant accident

International Nuclear Information System (INIS)

Zhang Guobin; Feng Jiaxuan

2012-01-01

From the nuclear accident in Japan Fuksuhima in March this year, despite a shut down of the reactor, the residue heat inside the reactor was not able to remove due to the failure of the cooling system and it finally caused the catastrophe. It was observed that when the failure of the cooling system after an earthquake of magnitude 9 and a tsunami of 28 meters height, the containment vessel for the reactor core was still able to maintain its integrity in the first 24 hours before the first explosion was happened. A backup emergency heat removal system for nuclear power plants using mo- bile pumps and liquid chilling units has been proposed 20 years ago by Cheung [Ref. 1]. Due to the fact that there are more than 400 nuclear power plants around the world and 10% of them are located in earthquake active zone, together with the aging of some of the power plants which were built more than 30 years ago, the risk of another nuclear accident becomes high. An emergency safety measure has to be designed in order to deal with the unforeseen scenario. This re- port re-visits the proposal again; to re-design to the suit the need and to integrate with the current situation of the nuclear industry. (authors)

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

International Nuclear Information System (INIS)

John Darrell Bess

2008-01-01

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for

Tokamak power reactor ignition and time dependent fractional power operation

International Nuclear Information System (INIS)

Vold, E.L.; Mau, T.K.; Conn, R.W.

1986-06-01

A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve

Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

International Nuclear Information System (INIS)

Berg, Thomas A.; Disney, Richard K.

2004-01-01

Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs

Reactor power cutback system test experience at YGN 4

International Nuclear Information System (INIS)

Chi, Sung Goo; Kim, Se Chang; Seo, Jong Tae; Eom, Young Meen; Wook, Jeong Dae; Choi, Young Boo

1995-01-01

YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

Physics and potentials of fissioning plasmas for space power and propulsion

Science.gov (United States)

Thom, K.; Schwenk, F. C.; Schneider, R. T.

1976-01-01

Fissioning uranium plasmas are the nuclear fuel in conceptual high-temperature gaseous-core reactors for advanced rocket propulsion in space. A gaseous-core nuclear rocket would be a thermal reactor in which an enriched uranium plasma at about 10,000 K is confined in a reflector-moderator cavity where it is nuclear critical and transfers its fission power to a confining propellant flow for the production of thrust at a specific impulse up to 5000 sec. With a thrust-to-engine weight ratio approaching unity, the gaseous-core nuclear rocket could provide for propulsion capabilities needed for manned missions to the nearby planets and for economical cislunar ferry services. Fueled with enriched uranium hexafluoride and operated at temperatures lower than needed for propulsion, the gaseous-core reactor scheme also offers significant benefits in applications for space and terrestrial power. They include high-efficiency power generation at low specific mass, the burnup of certain fission products and actinides, the breeding of U-233 from thorium with short doubling times, and improved convenience of fuel handling and processing in the gaseous phase.

10-75-kWe-reactor-powered organic Rankine-cycle electric power systems (ORCEPS) study. Final technical report

Energy Technology Data Exchange (ETDEWEB)

1977-03-30

This 10-75 kW(e) Reactor-ORCEPS study was concerned with the evaluation of several organic Rankine cycle energy conversion systems which utilized a /sup 235/U-ZrH reactor as a heat source. A liquid metal (NaK) loop employing a thermoelectric converter-powered EM pump was used to transfer the reactor energy to the organic working fluid. At moderate peak cycle temperatures (750/sup 0/F), power conversion unit cycle efficiencies of up to 25% and overall efficiencies of 20% can be obtained. The required operating life of seven years should be readily achievable. The CP-25 (toluene) working fluid cycle was found to provide the highest performance levels at the lowest system weights. Specific weights varies from 100 to 50 lb/kW(e) over the power level range 10 to 75 kW(e). (DLC)

Space nuclear reactor concepts for avoidance of a single point failure

International Nuclear Information System (INIS)

El-Genk, M. S.

2007-01-01

This paper presents three space nuclear reactor concepts for future exploration missions requiring electrical power of 10's to 100's kW, for 7-10 years. These concepts avoid a single point failure in reactor cooling; and they could be used with a host of energy conversion technologies. The first is lithium or sodium heat pipes cooled reactor. The heat pipes operate at a fraction of their prevailing capillary or sonic limit. Thus, when a number of heat pipes fail, those in the adjacent modules remove their heat load, maintaining reactor core adequately cooled. The second is a reactor with a circulating liquid metal coolant. The reactor core is divided into six identical sectors, each with a separate energy conversion loop. The sectors in the reactor core are neurotically coupled, but hydraulically decoupled. Thus, when a sector experiences a loss of coolant, the fission power generated in it will be removed by the circulating coolant in the adjacent sectors. In this case, however, the reactor fission power would have to decrease to avoid exceeding the design temperature limits in the sector with a failed loop. These two reactor concepts are used with energy conversion technologies, such as advanced Thermoelectric (TE), Free Piston Stirling Engines (FPSE), and Alkali Metal Thermal-to- Electric Conversion (AMTEC). Gas cooled reactors are a better choice to use with Closed Brayton Cycle engines, such as the third reactor concept to be presented in the paper. It has a sectored core that is cooled with a binary mixture of He-Xe (40 gm/mole). Each of the three sectors in the reactor has its own CBC and neutronically, but not hydraulically, coupled to the other sectors

Space nuclear reactor shields for manned and unmanned applications

International Nuclear Information System (INIS)

McKissock, B.I.; Bloomfield, H.S.

1990-01-01

Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. The shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station and advanced manned lunar base. (author)

Space nuclear reactor shields for manned and unmanned applications

International Nuclear Information System (INIS)

Mckissock, B.I.; Bloomfield, H.S.

1989-01-01

Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances, and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. Shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station, and advanced manned lunar base

10 KWe dual-mode space nuclear power system for military and scientific applications

International Nuclear Information System (INIS)

Malloy, J.; Westerman, K.; Rochow, R.; Scoles, S.

1992-01-01

This paper discusses a 10 KWe dual-mode space power system concept which has been identified and is based on INEL's Small Externally-fueled Heat Pipe Thermionic Reactor (SEHPTR) concept. This power system will enhance user capabilities by providing reliable electric power and by providing two propulsion systems; electric power for an arc-jet electric propulsion system and direct thrust by heating hydrogen propellant inside the reactor. The low thrust electric thrusters allow efficient station keeping and long-term maneuvering. This paper will focus on the nuclear power system design, including: the reactor with its UO 2 fuel in tungsten clad, 36 thermionic heat pipe modules (THPMs) which produce electricity within the reactor and remove waste heat, radiation shielding, waste heat radiators, and reactivity control systems. The use of non-vented fuel elements for short lifetime missions (under five years) will be described

14 CFR 25.1142 - Auxiliary power unit controls.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Auxiliary power unit controls. 25.1142... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Powerplant Controls and Accessories § 25.1142 Auxiliary power unit controls. Means must be provided on the flight deck for starting...

14 CFR 29.1142 - Auxiliary power unit controls.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Auxiliary power unit controls. 29.1142... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Powerplant Controls and Accessories § 29.1142 Auxiliary power unit controls. Means must be provided on the flight deck for starting...

Molten salt reactors and possible scenarios for future nuclear power deployment

International Nuclear Information System (INIS)

Merle-Lucotte, E.; Mathieu, L.; Heuer, D.; Loiseaux, J. M.; Billebaud, A.; Brissot, R.; David, S.; Garzenne, C.; Laulan, O.; Le Brun, C.; Lecarpentier, D.; Liatard, E.; Meplan, O.; Michel-Sendis, F.; Nuttin, A.; Perdu, F.

2004-01-01

An important fraction of the nature energy demand may be satisfied by nuclear power. In this context, the possibilities of worldwide nuclear deployment are studied. We are convinced that the Molten Salt Reactors may play a central role in this deployment. The Molten Salt Reactor needs to be coupled to a reprocessing unit in order to extract the Fission Products which poison the core. The efficiency of this reprocessing has a crucial influence on reactor behavior especially for the breeding ratio. The Molten Salt Breeder Reactor project was based on an intensive reprocessing for high breeding purposes. A new concept of Thorium Molten Salt Reactor is presented here. Including this new concept in the worldwide nuclear deployment, to satisfy these power needs, we consider three typical scenarios, based on three reactor types: Pressurized Water Reactor, Fast Neutron Reactor and Thorium Molten Salt Reactor. The aim of this paper is to demonstrate, in a first hand that a Thorium Molten Salt Reactor can be realistic, with correct temperature coefficients and at least iso-breeder with slow reprocessing and new geometry; on the other hand that such Molten Salt Reactors enable a successful nuclear deployment, while minimizing fuel and waste management problems. (authors)

CVD refractory metals and alloys for space nuclear power application

International Nuclear Information System (INIS)

Yang, L.; Gulden, T.D.; Watson, J.F.

1984-01-01

CVD technology has made significant contributions to the development of space nuclear power systems during the period 1962 to 1972. For the in-core thermionic concept, CVD technology is essential to the fabrication of the tungsten electron emitter. For the liquid metal cooled fuel pin using uranium nitride as fuel and T-111 and Nb-1 Zr as cladding, a tungsten barrier possibly produced by CVD methods is essential to the fuel-cladding compatibility at the designed operating temperature. Space power reactors may use heat pipes to transfer heat from the reactor core to the conversion system. CVD technology has been used for fabricating the heat pipe used as cross-flow heat exchanger, including the built-in channels on the condenser wall for liquid lithium return. 28 references, 17 figures

Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

International Nuclear Information System (INIS)

Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

2008-01-01

The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

Floating nuclear heat. And power station 'Pevec' with KLT-40S type reactor plant for remote regions of Russia

International Nuclear Information System (INIS)

Veshnyakov, K.B.; Kiryushin, A.I.; Panov, Yu.K.; Polunichev, V.I.

2000-01-01

Floating small nuclear power plants power for local energy systems of littoral regions of Russia, located far from central energy system, open a new line in nuclear power development. Designing a floating power unit of a lead nuclear heat and power generating station for port Pevec at the Chuckchee national district is currently nearing completion. Most labor-intensive components are being manufactured. The co-generation NPP Pevec is to be created on the basis of a floating power unit with KLT-40S type reactor plant. KLT-40S reactor plant is based on similar propulsion plants, verified at operation of Russia's nuclear-powered civil ships, evolutionary improved by elimination of 'weak points' revealed during its prototypes operation or on the basis of safety analysis. KLT-40S reactor plant uses the most wide-spread and developed in the world practice PWR-type reactor. KLT-40S meets contemporary national and international requirements imposed to future reactor plants. The NHPS description, its main technical-economic data, environmental safety indices, basic characteristics of KLT-40S reactor plant are presented. Prospects of small NPPs utilization outside Russia, particularly as an energy source for sea water desalination, are considered. (author)

Ultrasonic meters in the feedwater flow to recover thermal power in the reactor of nuclear power plant of Laguna Verde U1 and U2

International Nuclear Information System (INIS)

Tijerina S, F.

2008-01-01

The engineers in nuclear power plants BWRs and PWRs based on the development of the ultrasonic technology for the measurement of the mass, volumetric flow, density and temperature in fluids, have applied this technology in two primary targets approved by the NRC: the use for the recovery of thermal power in the reactor and/or to be able to realize an increase of thermal power licensed in a 2% (MUR) by 1OCFR50 Appendix K. The present article mentions the current problem in the measurement of the feedwater flow with Venturi meters, which affects that the thermal balance of reactor BWRs or PWRs this underestimated. One in broad strokes describes the application of the ultrasonic technology for the ultrasonic measurement in the flow of the feedwater system of the reactor and power to recover thermal power of the reactor. One is to the methodology developed in CFE for a calibration of the temperature transmitters of RTD's and the methodology for a calibration of the venturi flow transmitters using ultrasonic measurement. Are show the measurements in the feedwater of reactor of the temperature with RTD's and ultrasonic measurement, as well as the flow with the venturi and the ultrasonic measurement operating the reactor to the 100% of nominal thermal power, before and after the calibration of the temperature transmitters and flow. Finally, is a plan to be able to realize a recovery of thermal power of the reactor, showing as carrying out their estimations. As a result of the application of ultrasonic technology in the feedwater of reactor BWR-5 in Laguna Verde, in the Unit 1 cycle 13 it was recover an equivalent energy to a thermal power of 25 MWt in the reactor and an exit electrical power of 6 M We in the turbogenerator. Also in the Unit 2 cycle 10 it was recover an equivalent energy to a thermal power of 40 MWt in the reactor and an exit electrical power of 16 M We in the turbogenerator. (Author)

Safstor decommissioning of the Humboldt Bay Power Plant Unit No. 3

International Nuclear Information System (INIS)

Nelson, R.T.

1985-01-01

The Humboldt Bay Power Plant is located near Eureka, California, about 265 miles north of San Francisco. The plant consists of two fossil fueled units, two mobil gas turbine peaking units, and a nuclear unit - Unit No. 3. Unit No. 3, which utilized a boiling water reactor, was constructed between 1960 and 1963. The unit began commercial operation in August 1963 and operated until July 2, 1976 when it was shutdown for refueling, seismic modifications, and additional seismic and geologic studies. During the years Unit 3 operated it had one of the best operating records of any nuclear power plant in the United States. For its operating lifetime Unit 3 had an overall capacity factor of 63.0% and an availability factor of 85.9%. The unit included certain design features which made it unique among nuclear power plants of its era. Some of these unique features included natural circulation recirculation flow which eliminated the need for costly recirculation pumps, utilization of a pressure suppression containment system which had been developed jointly by PG and E and the General Electric Company, and the fact that the reactor vessel and the containment system were constructed in a caisson below ground level. These design features reduced the overall construction cost of the unit and improved its inherent safety

Determination of noise sources and space-dependent reactor transfer functions from measured output signals only

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E.; van Dam, H.; Kleiss, E.B.J.; van Uitert, G.C.; Veldhuis, D.

1982-01-01

The measured cross power spectral densities of the signals from three neutron detectors and the displacement of the control rod of the 2 MW research reactor HOR at Delft have been used to determine the space-dependent reactor transfer function, the transfer function of the automatic reactor control system and the noise sources influencing the measured signals. From a block diagram of the reactor with control system and noise sources expressions were derived for the measured cross power spectral densities, which were adjusted to satisfy the requirements following from the adopted model. Then for each frequency point the required transfer functions and noise sources could be derived. The results are in agreement with those of autoregressive modelling of the reactor control feed-back loop. A method has been developed to determine the non-linear characteristics of the automatic reactor control system by analysing the non-gaussian probability density function of the power fluctuations.

Determination of noise sources and space-dependent reactor transfer functions from measured output signals only

International Nuclear Information System (INIS)

Hoogenboom, J.E.

1982-01-01

The measured cross power spectral densities of the signals from three neutron detectors and the displacement of the control rod of the 2 MW research reactor HOR at Delft have been used to determine the space-dependent reactor transfer function, the transfer function of the automatic reactor control system and the noise sources influencing the measured signals. From a block diagram of the reactor with control system and noise sources expressions were derived for the measured cross power spectral densities, which were adjusted to satisfy the requirements following from the adopted model. Then for each frequency point the required transfer functions and noise sources could be derived. The results are in agreement with those of autoregressive modelling of the reactor control feed-back loop. A method has been developed to determine the non-linear characteristics of the automatic reactor control system by analysing the non-gaussian probability density function of the power fluctuations. (author)

Reactor power control device

International Nuclear Information System (INIS)

Kobayashi, Akira.

1980-01-01

Purpose: To prevent misoperation in a control system for the adjustment of core coolant flow rate, and the increase in the neutron flux density caused from the misoperation in BWR type reactors. Constitution: In a reactor power control system adapted to control the reactor power by the adjustment of core flow rate, average neutron flux signals of a reactor core, entire core flow rate signals and operation state signals for coolant recycling system are inputted to a microcomputer. The outputs from the computer are sent to a recycling MG set speed controller to control the reactor core flow rate. The computer calculates the change ratio with time in the average neutron flux signals, correlation between the average neutron flux signals and the entire core flow rate signals, change ratio with time in the operation state signals for the coolant recycling system and the like and judges the abnormality in the coolant recycling system based on the calculated results. (Ikeda, J.)

Upgrade of KNPEC no.2 Simulator for Kori Unit 3 Power Uprating

Energy Technology Data Exchange (ETDEWEB)

Hong, Jin-Hyuk; Lee, Seung-Ho [KEPRI, Daejeon (Korea, Republic of)

2007-07-01

Kori-Unit 3 and 4 is preparing the operation of the power-uprating (2900MWt), and therefore the Korea regulatory body(KINS) requested the operator training with the simulator reflecting the power-uprating. As a result of the intensive research and expertise of KEPRI on the simulators, KEPRI accomplished the upgrade project of KNPEC no.2 simulator for Kori-Unit 3 power-uprating. This project includes various high-tech methods incorporating - realtime neutronics model based on MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors) code, best-estimate neutronics code by the KINS, (By using the RMASTER, the precision of the simulation of the neutron behaviors in the core is highly improved.) - betterment of the reactor coolant system and the balance-of-plant system - modification of the corresponding setpoints due to the power-uprating And the acceptance test procedure (ATP) was successfully carried out through the integration of system models and its performance tests. Through the success of this project, the operator training for the power uprating of the Kori-Unit 3 will be accomplished before its power operation and, after all, this simulator will contribute to the safe operation for the power-uprating of the Kori-Unit 3 and 4.

Nuclear power desalinating complex with IRIS reactor plant and Russian distillation desalinating unit

International Nuclear Information System (INIS)

Kostin, V. I.; Panov, Yu.K.; Polunichev, V. I.; Fateev, S. A.; Gureeva, L. V.

2004-01-01

This paper has been prepared as a result of Russian activities on the development of nuclear power desalinating complex (NPDC) with the IRIS reactor plant (RP). The purpose of the activities was to develop the conceptual design of power desalinating complex (PDC) and to evaluate technical and economical indices, commercial attractiveness and economical efficiency of PDC based on an IRIS RP with distillation desalinating plants. The paper presents the main results of studies as applied to dual-purpose PDC based on IRIS RP with different types of desalinating plants, namely: characteristics of nuclear power desalinating complex based on IRIS reactor plant using Russian distillation desalinating technologies; prospective options of interface circuits of the IRIS RP with desalinating plants; evaluations of NPDC with IRIS RP output based on selected desalinating technologies for water and electric power supplied to the grid; cost of water generated by NPDC for selected interface circuits made by the IAEA DEEP code as well as by the Russian TEO-INVEST code; cost evaluation results for desalinated water of PDC operating on fossil fuel and conditions for competitiveness of the nuclear PDC based on IRIS RP compared with analog desalinating complexes operating on fossil fuel.(author)

Green Application for Space Power

Science.gov (United States)

Robinson, Joel

2015-01-01

Most space vehicle auxiliary power units (APUs) use hydrazine propellant for generating power. Hydrazine is a toxic, hazardous fuel that requires special safety equipment and processes for handling and loading. In recent years, there has been development of two green propellants (less toxic) that could enable their use in APUs. The Swedish government, in concert with the Swedish Space Corporation, has developed a propellant based on ammonium dinitramide (LMP-103S) that was flown on the Prisma spacecraft in 2010. The United States Air Force (USAF) has been developing a propellant based on hydroxylammonium nitrate (AFM315E) that is scheduled to fly on the Green Propellant Infusion Mission in the spring of 2016 to demonstrate apogee and reaction control thrusters. However, no one else in the Agency is currently pursuing use of green propellants for application to the APUs. Per the TA-01 Launch Propulsion Roadmap, the Space Technology Mission Directorate had identified the need to have a green propellant APU by 2015. This is our motivation for continuing activities.

Evaluation on radioactive waste disposal amount of Kori Unit 1 reactor vessel considering cutting and packaging methods

International Nuclear Information System (INIS)

Choi, Yu Jong; Lee, Seong Cheol; Kim, Chang Lak

2016-01-01

Decommissioning of nuclear power plants has become a big issue in South Korea as some of the nuclear power plants in operation including Kori unit 1 and Wolsung unit 1 are getting old. Recently, Wolsung unit 1 received permission to continue operation while Kori unit 1 will shut down permanently in June 2017. With the consideration of segmentation method and disposal containers, this paper evaluated final disposal amount of radioactive waste generated from decommissioning of the reactor pressure vessel in Kori unit 1 which will be decommissioned as the first in South Korea. The evaluation results indicated that the final disposal amount from the top and bottom heads of the reactor pressure vessel with hemisphere shape decreased as they were cut in smaller more effectively than the cylindrical part of the reactor pressure vessel. It was also investigated that 200 L and 320 L radioactive waste disposal containers used in Kyung-Ju disposal facility had low payload efficiency because of loading weight limitation

Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications