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Sample records for candu power reactor

  1. Nuclear Archeology for CANDU Power Reactors

    SciTech Connect

    Broadhead, Bryan L

    2011-01-01

    The goal of this work is the development of so-called 'nuclear archeology' techniques to predict the irradiation history of both fuel-related and non-fuel-related materials irradiated in the CANDU (CANada Deuterium Uranium) family of nuclear reactors. In this application to CANDU-type reactors, two different scenarios for the collection of the appropriate data for use in these procedures will be assumed: the first scenario is the removal of the pressure tubes, calandria tubes, or fuel cladding and destructive analysis of the activation products contained in these structural materials; the second scenario is the nondestructive analysis (NDA) of the same hardware items via high-resolution gamma ray scans. There are obvious advantages and disadvantages for each approach; however, the NDA approach is the central focus of this work because of its simplicity and lack of invasiveness. The use of these techniques along with a previously developed inverse capability is expected to allow for the prediction of average flux levels and irradiation time, and the total fluence for samples where the values of selected isotopes can be measured.

  2. Tritium activities in Canada supporting CANDU{sup R} nuclear power reactors

    SciTech Connect

    Miller, J. M.

    2008-07-15

    An overview of the various Canadian tritium research and operational activities supporting the development, refurbishment and operation of CANDU{sup R} nuclear power reactors is presented. These activities encompass tritium health and safety, tritium in the environment, tritium interaction with materials, and tritium processing, and relate to both supporting R and D advances as well as operational best practices. The collective results of these activities contribute to our goals of improving worker and public safety, and operational efficiency. (authors)

  3. The CANDU Reactor System: An Appropriate Technology.

    PubMed

    Robertson, J A

    1978-02-10

    CANDU power reactors are characterized by the combination of heavy water as moderator and pressure tubes to contain the fuel and coolant. Their excellent neutron economy provides the simplicity and low costs of once-through natural-uranium fueling. Future benefits include the prospect of a near-breeder thorium fuel cycle to provide security of fuel supply without the need to develop a new reactor such as the fast breeder. These and other features make the CANDU system an appropriate technology for countries, like Canada, of intermediate economic and industrial capacity. PMID:17788102

  4. Actinide Burning in CANDU Reactors

    SciTech Connect

    Hyland, B.; Dyck, G.R.

    2007-07-01

    Actinide burning in CANDU reactors has been studied as a method of reducing the actinide content of spent nuclear fuel from light water reactors, and thereby decreasing the associated long term decay heat load. In this work simulations were performed of actinides mixed with natural uranium to form a mixed oxide (MOX) fuel, and also mixed with silicon carbide to form an inert matrix (IMF) fuel. Both of these fuels were taken to a higher burnup than has previously been studied. The total transuranic element destruction calculated was 40% for the MOX fuel and 71% for the IMF. (authors)

  5. Thermochemical modelling of advanced CANDU reactor fuel

    NASA Astrophysics Data System (ADS)

    Corcoran, Emily Catherine

    2009-04-01

    With an aging fleet of nuclear generating facilities, the imperative to limit the use of non-renewal fossil fuels and the inevitable need for additional electricity to power Canada's economy, a renaissance in the use of nuclear technology in Canada is at hand. The experience and knowledge of over 40 years of CANDU research, development and operation in Ontario and elsewhere has been applied to a new generation of CANDU, the Advanced CANDU Reactor (ACR). Improved fuel design allows for an extended burnup, which is a significant improvement, enhancing the safety and the economies of the ACR. The use of a Burnable Neutron Absorber (BNA) material and Low Enriched Uranium (LEU) fuel has created a need to understand better these novel materials and fuel types. This thesis documents a work to advance the scientific and technological knowledge of the ACR fuel design with respect to thermodynamic phase stability and fuel oxidation modelling. For the BNA material, a new (BNA) model is created based on the fundamental first principles of Gibbs energy minimization applied to material phase stability. For LEU fuel, the methodology used for the BNA model is applied to the oxidation of irradiated fuel. The pertinent knowledge base for uranium, oxygen and the major fission products is reviewed, updated and integrated to create a model that is applicable to current and future CANDU fuel designs. As part of this thesis, X-Ray Diffraction (XRD) and Coulombic Titration (CT) experiments are compared to the BNA and LEU models, respectively. From the analysis of the CT results, a number of improvements are proposed to enhance the LEU model and provide confidence in its application to ACR fuel. A number of applications for the potential use of these models are proposed and discussed. Keywords: CANDU Fuel, Gibbs Energy Mimimization, Low Enriched Uranium (LEU) Fuel, Burnable Neutron Absorber (BNA) Material, Coulometric Titration, X-Ray Diffraction

  6. Systems analysis of the CANDU 3 Reactor

    SciTech Connect

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H.

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  7. Optimization of the Mode of the Uranium-233 Accumulation for Application in Thorium Self-Sufficient Fuel Cycle of Candu Power Reactor

    SciTech Connect

    Bergelson, Boris; Gerasimov, Alexander; Tikhomirov, Georgy

    2006-07-01

    Results of calculation studies of the first stage of self-sufficient thorium cycle for CANDU reactor are presented in the paper. The first stage is preliminary accumulation of {sup 233}U in the CANDU reactor itself. Parameters of active core and scheme of fuel reloading were accepted the same as those for CANDU reactor. It was assumed for calculations, that enriched {sup 235}U or plutonium was used as additional fissile material to provide neutrons for {sup 233}U production. Parameters of 10 different variants of the elementary cell of active core were calculated for the lattice pitch, geometry of fuel channels, and fuel assembly of the CANDU reactor. The results presented in the paper allow to determine the time of accumulation of the required amount of {sup 233}U and corresponding number of targets going into processing for {sup 233}U extraction. Optimum ratio of the accumulation time to number of processed targets can be determined using the cost of electric power produced by the reactor and cost of targets along with their processing. (authors)

  8. Thermal-hydraulic interfacing code modules for CANDU reactors

    SciTech Connect

    Liu, W.S.; Gold, M.; Sills, H.

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  9. A CANDU-Based Fast Irradiation Reactor

    SciTech Connect

    Shatilla, Youssef

    2006-07-01

    A new steady-state fast neutron reactor is needed to satisfy the testing needs of Generation IV reactors, the Space Propulsion Program, and the Advanced Fuel Cycle Initiative. This paper presents a new concept for a CANDU-based fast irradiation reactor that is horizontal in orientation, with individual pressure tubes running the entire length of the scattering-medium tank (Calandria) filled with Lead-Bismuth-Eutectic (LBE). This approach for a test reactor will provide more flexibility in refueling, sample removal, and ability to completely re-configure the core to meet different users' requirements. Full core neutronic analysis of several fuel/coolant/geometry combinations showed a small hexagonal, LBE-cooled, U-Pu-10Zr fuel, with a core power of 100 MW{sub th} produced a fast flux (>0.1 MeV) of 1.5 x 10{sup 15} n/cm{sup 2} sec averaged over the whole length of six irradiation channels with a total testing volume of more than 77 liters. In-core breeding allowed the Pu-239 enrichment to be 15.3% which should result in core continuous operation for 180 effective full power days. Other coolants investigated included high pressure water steam and helium. An innovative shutdown/control system which consisted of the six outermost fuel channels was proven to be effective in shutting the core down when flooded with boric acid as a neutron absorber. The new shutdown/control system has the advantage of causing the minimum perturbation of the axial flux shape when the control channels are partially flooded with boric acid. This is because the acid is injected homogeneously along the control channel in contrast to regular control rods that are injected partially causing an axial perturbation in the core flux which in turn reduces safety analysis margins. The new shutdown/control system is not required to penetrate the core in a direction vertical to the fuel channels which allowed the freedom of changing core pitch as deemed necessary. A preliminary thermal hydraulic analysis

  10. Leak detection capability in CANDU reactors

    SciTech Connect

    Azer, N.; Barber, D.H.; Boucher, P.J.

    1997-04-01

    This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis.

  11. CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

    SciTech Connect

    Charak, I.; Kier, P.H.

    1995-04-01

    Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule ({section}50.62); (2) station blackout ({section}50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term ({section}50.34(f) and {section}100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements ({section}50.55a, etc); (6) ECCS acceptance criteria ({section}50.46)(b); (7) combustible gas control ({section}50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle ({section}51.51); and (11) (standards {section}50.55a).

  12. Recommendations for damping and treatment of modeling uncertainty in seismic analysis of CANDU nuclear power plant

    SciTech Connect

    Usmani, S.A.; Baughman, P.D.

    1996-12-01

    The seismic analysis of the CANDU nuclear power plant is governed by Canadian Standard series N289. However, the dynamic analysis of some equipment and system such as the CANDU reactor and fueling machine must treat unique components not directly covered by the broad recommendations of these standards. This paper looks at the damping values and treatment of modeling uncertainty recommended by CSA N289.3, the current state of knowledge and expert opinion as reflected in several current standards, testing results, and the unique aspects of the CANDU system. Damping values are recommended for the component parts of the CANDU reactor and fueling machine system: reactor building, calandria vault, calandria, fuel channel, pressure tube, fueling machine and support structure. Recommendations for treatment of modeling and other uncertainties are also presented.

  13. CANDU in-reactor quantitative visual-based inspection techniques

    NASA Astrophysics Data System (ADS)

    Rochefort, P. A.

    2009-02-01

    This paper describes two separate visual-based inspection procedures used at CANDU nuclear power generating stations. The techniques are quantitative in nature and are delivered and operated in highly radioactive environments with access that is restrictive, and in one case is submerged. Visual-based inspections at stations are typically qualitative in nature. For example a video system will be used to search for a missing component, inspect for a broken fixture, or locate areas of excessive corrosion in a pipe. In contrast, the methods described here are used to measure characteristic component dimensions that in one case ensure ongoing safe operation of the reactor and in the other support reactor refurbishment. CANDU reactors are Pressurized Heavy Water Reactors (PHWR). The reactor vessel is a horizontal cylindrical low-pressure calandria tank approximately 6 m in diameter and length, containing heavy water as a neutron moderator. Inside the calandria, 380 horizontal fuel channels (FC) are supported at each end by integral end-shields. Each FC holds 12 fuel bundles. The heavy water primary heat transport water flows through the FC pressure tube, removing the heat from the fuel bundles and delivering it to the steam generator. The general design of the reactor governs both the type of measurements that are required and the methods to perform the measurements. The first inspection procedure is a method to remotely measure the gap between FC and other in-core horizontal components. The technique involves delivering vertically a module with a high-radiation-resistant camera and lighting into the core of a shutdown but fuelled reactor. The measurement is done using a line-of-sight technique between the components. Compensation for image perspective and viewing elevation to the measurement is required. The second inspection procedure measures flaws within the reactor's end shield FC calandria tube rolled joint area. The FC calandria tube (the outer shell of the FC) is

  14. Single channel flow blockage accident phenomena identification and ranking table (PIRT) for the advanced Candu reactor

    SciTech Connect

    Popov, N.K.; Abdul-Razzak, A.; Snell, V.G.; Langman, V.; Sills, H.

    2004-07-01

    The Advanced Candu Reactor (ACRTM) is an evolutionary advancement of the current Candu 6{sup R} reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by a heavy water moderator, as with all Candu reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper documents the results of Phenomena Identification and Ranking Table (PIRT) results for a very limited frequency, beyond design basis event of the ACR design. This PIRT is developed in a highly structured process of expert elicitation that is well supported by experimental data and analytical results. The single-channel flow blockage event in an ACR reactor assumes a severe flow blockage of one of the reactor fuel channels, which leads to a reduction of the flow in the affected channel, leading to fuel cladding and fuel temperature increase. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the flow blockage phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the finalized PIRT tables. (authors)

  15. Plutonium Consumption Program, CANDU Reactor Project final report

    SciTech Connect

    Not Available

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  16. NRC review of the CANDU-3 reactor design

    SciTech Connect

    Kennedy, J.L. )

    1993-01-01

    This paper presents an overview of the US Nuclear Regulatory Commission's (NRC's) effort to complete an early review of the Canada deuterium uranium (CANDU)-3 reactor design prior to formal submittal of an application for standard design certification. The NRC is conducting a review of the CANDU-3 design in support of a request by AECL Technologics, the US sponsor of the design. The purpose of this review is to encourage early interactions by applicants, vendors, and government agencies with the NRC. The CANDU-3 design is being developed by Atomic Energy of Canada, Limited (AECL), whose CANDU operations are based in Mississauga, Ontario. AECL Technologies, a US subsidiary of AECL, Incorporated, informed the NRC of its intent to seek design certification of the CANDU-3 design under the provisions of 10CFR52 in a letter to the NRC dated May 25, 1989. This paper describes the commission's basis for this type of early review, its purposes and objectives, key elements of the review, the intended product, and the schedule.

  17. An on-line regional overpower surveillance system for Candu reactors

    SciTech Connect

    Wallace, D. J.; Caxaj, V.; Seidu, A. S.; Hartmann, W.; Sur, B.; McDonald, A.

    2006-07-01

    The current methodology for establishing Regional Overpower Protection (ROP) trip set-points for Canada Deuterium Uranium (Candu{sup R} reactors requires an extensive and detailed assessment of the plant based on a distribution of channel and bundle powers (flux shapes) calculated from a range of device configurations (e.g., zone controller levels, adjuster bank movements, mechanical control absorber movements, shut-off rod insertions) and a set of thermalhydraulic plant data (channel flows, reactor inlet-header temperatures, channel differential pressure). An on-line approach would provide an interface to assist operators in routine monitoring, diagnostic and maintenance activities by providing Critical Channel Powers (CCP) and ROP set points from instantaneous flux shapes derived from real-time detector readings and associated thermalhydraulic conditions. This paper describes an Advanced On-Line Regional Overpower Surveillance (AOL-ROS) system currently under development at Atomic Energy of Canada Limited (AECL) for Candu reactors. Development has been based on an assessment using instantaneous operating data for the period February to April 2004 from a Candu 6 reactor located at Point Lepreau, New Brunswick (Canada). (authors)

  18. Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond

    SciTech Connect

    Ovanes, M.; Chan, P. S. W.; Mao, J.; Alderson, N.; Hopwood, J. M.

    2012-07-01

    The Enhanced CANDU 6{sup R} (ECo{sup R}) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

  19. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    SciTech Connect

    Greene, S.R.; Spellman, D.J.; Bevard, B.B.

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  20. Modelling the activity of 129I in the primary coolant of a CANDU reactor

    NASA Astrophysics Data System (ADS)

    Lewis, Brent J.; Husain, Aamir

    2003-01-01

    A mathematical treatment has been developed to describe the activity levels of 129I as a function of time in the primary heat transport system during constant power operation and for a reactor shutdown situation. The model accounts for a release of fission-product iodine from defective fuel rods and tramp uranium contamination on in-core surfaces. The physical transport constants of the model are derived from a coolant activity analysis of the short-lived radioiodine species. An estimate of 3×10 -9 has been determined for the coolant activity ratio of 129I/ 131I in a CANDU Nuclear Generating Station (NGS), which is in reasonable agreement with that observed in the primary coolant and for plant test resin columns from pressurized and boiling water reactor plants. The model has been further applied to a CANDU NGS, by fitting it to the observed short-lived iodine and long-lived cesium data, to yield a coolant activity ratio of ˜2×10 -8 for 129I/ 137Cs. This ratio can be used to estimate the levels of 129I in reactor waste based on a measurement of the activity of 137Cs.

  1. Transmutation of actinides in power reactors.

    PubMed

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides. PMID:16604724

  2. The application of Plant Reliability Data Information System (PRINS) to CANDU reactor

    SciTech Connect

    Hwang, S. W.; Lim, Y. H.; Park, H. C.

    2012-07-01

    As risk-informed applications (RIAs) are actively implanted in the nuclear industry, an issue associated with technical adequacy of Probabilistic Safety Assessment (PSA) arises in its modeling and data sourcing. In Korea, PSA for all Korean NPPs has been completed and KHNP(Korea Hydro and Nuclear Power Plant Company) developed the database called the Plant Reliability Data Information System (PRinS). It has several characteristics that distinguish it from other database system such as NPRDs (INPO,1994), PRIS (IAEA), and SRDF (EdF). This database has the function of systematic data management such as automatic data-gathering, periodic data deposition and updating, statistical analysis including Bayesian method, and trend analysis of failure rate or unavailability. In recent PSA for CANDU reactor, the component failure data of EPRI ALWR URD and Component Reliability Database were preferentially used as generic data set. The error factor for most component failure data was estimated by using the information NUREG/CR-4550 and NUREG/CR-4639. Also, annual trend analysis was performed for the functional losses of components using the statistical analysis and chart module of PRinS. Furthermore, the database has been updated regularly and maintained as a living program to reflect the current status. This paper presents the failure data analysis using PRinS which provides Bayesian analysis on main components in the CANDU reactor. (authors)

  3. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    SciTech Connect

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L.

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  4. A dynamic fuel cycle analysis for a heterogeneous thorium-DUPIC recycle in CANDU reactors

    SciTech Connect

    Jeong, C. J.; Park, C. J.; Choi, H.

    2006-07-01

    A heterogeneous thorium fuel recycle scenario in a Canada deuterium uranium (CANDU) reactor has been analyzed by the dynamic analysis method. The thorium recycling is performed through a dry process which has a strong proliferation resistance. In the fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides, and fission products of a multiple thorium recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. The analysis results have shown that the heterogeneous thorium fuel cycle can be constructed through the dry process technology. It is also shown that the heterogeneous thorium fuel cycle can reduce the spent fuel inventory and save on the natural uranium resources when compared with the once-through cycle. (authors)

  5. CANDU-PHW reactor in relation to a Second Nuclear Era

    SciTech Connect

    Siddall, E.

    1984-05-01

    The reasons for the emergence and successful development of the CANDU-PHWR system are outlined. A plant of typical present-day design is described, with tabulated parameters and dimensions for two sizes. Technical, economic, and commercial differences from other reactor types are considered, including investment patterns and consumption of the basic uranium resource. Safety aspects are compared with those of other reactor types and a condensed risk assessment is given. The separate question of licensability is discussed, with the conclusion that this is unpredictable because of the nature of the licensing process. It is concluded that CANDU-PHWR is a viable system for a Second Nuclear Era.

  6. POWER REACTOR

    DOEpatents

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

  7. Once-through CANDU reactor models for the ORIGEN2 computer code

    SciTech Connect

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given.

  8. Materials performance in CANDU reactors: The first 30 years and the prognosis for life extension and new designs

    NASA Astrophysics Data System (ADS)

    Tapping, R. L.

    2008-12-01

    A number of CANDU reactors have now been in-service for more than 30 years, and several are planning life extensions. This paper summarizes the major corrosion degradation operating experience of various out-of-core (i.e., excluding fuel channels and fuel) materials in-service in currently operating CANDU reactors. Also discussed are the decisions that need to be made for life extension of replaceable and non-replaceable components such as feeders and steam generators, and materials choices for new designs, such as the advanced CANDU reactor (ACR) and enhanced CANDU-6. The basis for these choices, including a brief summary of the R&D necessary to support such decisions is provided. Finally we briefly discuss the materials and R&D needs beyond the immediate future, including new concepts to improve plant operability and component reliability.

  9. Plutonium Consumption Program, CANDU Reactor Project: Feasibility of BNFP Site as MOX Fuel Supply Facility. Final report

    SciTech Connect

    1995-06-30

    An evaluation was made of the technical feasibility, cost, and schedule for converting the existing unused Barnwell Nuclear Fuel Facility (BNFP) into a Mixed Oxide (MOX) CANDU fuel fabrication plant for disposition of excess weapons plutonium. This MOX fuel would be transported to Ontario where it would generate electricity in the Bruce CANDU reactors. Because CANDU MOX fuel operates at lower thermal load than natural uranium fuel, the MOX program can be licensed by AECB within 4.5 years, and actual Pu disposition in the Bruce reactors can begin in 2001. Ontario Hydro will have to be involved in the entire program. Cost is compared between BNFP and FMEF at Hanford for converting to a CANDU MOX facility.

  10. The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor

    SciTech Connect

    Morreale, A. C.; Ball, M. R.; Novog, D. R.; Luxat, J. C.

    2012-07-01

    The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a CANDU{sup R}* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)

  11. Optimization and implementation study of plutonium disposition using existing CANDU Reactors. Final report

    SciTech Connect

    1996-09-01

    Since early 1994, the Department of Energy has been sponsoring studies aimed at evaluating the merits of disposing of surplus US weapons plutonium as Mixed Oxide (MOX) fuel in existing commercial Canadian Pressurized Heavy Water reactors, known as CANDU`s. The first report, submitted to DOE in July, 1994 (the 1994 Executive Summary is attached), identified practical and safe options for the consumption of 50 to 100 tons of plutonium in 25 years in some of the existing CANDU reactors operating the Bruce A generating station, on Lake Huron, about 300 km north east of Detroit. By designing the fuel and nuclear performance to operate within existing experience and operating/performance envelope, and by utilizing existing fuel fabrication and transportation facilities and methods, a low cost, low risk method for long term plutonium disposition was developed. In December, 1995, in response to evolving Mission Requirements, the DOE requested a further study of the CANDU option with emphasis on more rapid disposition of the plutonium, and retaining the early start and low risk features of the earlier work. This report is the result of that additional work.

  12. The Advanced Candu reactor annunciation system - Compliance with IEC standard and US NRC guidelines

    SciTech Connect

    Leger, R.; Malcolm, S.; Davey, E.

    2006-07-01

    Annunciation is a key plant information system that alerts Operations staff to important changes in plant processes and systems. Operational experience at nuclear stations worldwide has shown that many annunciation implementations do not provide the support needed by Operations staff in all plant situations. To address utility needs for annunciation improvement in Candu plants, AECL in partnership with Canadian Candu utilities, undertook an annunciation improvement program in the early nineties. The outcome of the research and engineering development program was the development and simulator validation of alarm processing, display, and information presentation techniques that provide practical and effective solutions to key operational deficiencies with earlier annunciation implementations. The improved annunciation capabilities consist of a series of detection, information processing and presentation functions called the Candu Annunciation Message List System (CAMLS). The CAMLS concepts embody alarm processing, presentation and interaction techniques, and strategies and methods for annunciation system configuration to ensure improved annunciation support for all plant situations, especially in upset situations where the alarm generation rate is high. The Advanced Candu Reactor (ACR) project will employ the CAMLS annunciation concepts as the basis for primary annunciation implementations. The primary annunciation systems will be implemented from CAMLS applications hosted on AECL Advanced Control Centre Information System (ACCIS) computing technology. The ACR project has also chosen to implement main control room annunciation aspects in conformance with the following international standard and regulatory review guide for control room annunciation practice: International Electrotechnical Commission (IEC) 62241 - Main Control Room, Alarm Function and Presentation (International standard) US NRC NUREG-0700 - Human-System Interface Design Review Guidelines, Section 4

  13. NEUTRONIC REACTOR POWER PLANT

    DOEpatents

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

  14. Operating US power reactors

    SciTech Connect

    Silver, E.G.

    1982-07-01

    The operation of US power reactors during March and April 1982 is summarized. Events of special note are discussed in the text, and the operational performance of all licensed power reactors is presented. These data are taken from the monthly Operating Units Status Report prepared by the Nuclear Regulatory Commission (NRC).

  15. Implementation of an on-line monitoring system for transmitters in a CANDU nuclear power plant

    NASA Astrophysics Data System (ADS)

    Labbe, A.; Abdul-Nour, G.; Vaillancourt, R.; Komljenovic, D.

    2012-05-01

    Many transmitters (pressure, level and flow) are used in a nuclear power plant. It is necessary to calibrate them periodically to ensure that their measurements are accurate. These calibration tasks are time consuming and often contribute to worker radiation exposure. Human errors can also sometimes degrade their performance since the calibration involves intrusive techniques. More importantly, experience has shown that the majority of current calibration efforts are not necessary. These facts motivated the nuclear industry to develop new technologies for identifying drifting instruments. These technologies, well known as on-line monitoring (OLM) techniques, are non-intrusive and allow focusing the maintenance efforts on the instruments that really need a calibration. Although few OLM systems have been implemented in some PWR and BWR plants, these technologies are not commonly used and have not been permanently implemented in a CANDU plant. This paper presents the results of a research project that has been performed in a CANDU plant in order to validate the implementation of an OLM system. An application project, based on the ICMP algorithm developed by EPRI, has been carried out in order to evaluate the performance of an OLM system. The results demonstrated that the OLM system was able to detect the drift of an instrument in the majority of the studied cases. A feasibility study has also been completed and has demonstrated that the implementation of an OLM system at a CANDU nuclear power plant could be advantageous under certain conditions.

  16. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOEpatents

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  17. Compact power reactor

    DOEpatents

    Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

  18. Pressurized heavy water reactor fuel behaviour in power ramp conditions

    NASA Astrophysics Data System (ADS)

    Ionescu, S.; Uţă, O.; Pârvan, M.; Ohâi, D.

    2009-03-01

    In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.

  19. Outdoor field evaluation of passive tritiated water vapor samplers at Canadian power reactor sites.

    PubMed

    Wood, M J

    1996-02-01

    Tritium is one of several radioactive nuclides routinely monitored in and around CANDU (CANada Deuterium Uranium) power reactor facilities. Over the last ten years, passive samplers have replaced active sampling devices for sampling tritiated water vapor in the workplace at many CANDU stations. The potential of passive samplers for outdoor monitoring has also been realized. This paper presents the results of a 1-y field trial carried out at all five Canadian CANDU reactor sites. The results indicate that passive samplers can be used at most sampling locations to measure tritiated water vapor in air concentrations as low as 1 Bq m-3 over a 30-d sampling period. Only in one of the five sampling locations was poor agreement observed between active and passive monitoring data. This location, however, was very windy and it is suspected that the gusty winds were the source of the discrepancies observed. PMID:8567295

  20. Monte Carlo estimation of the dose and heating of cobalt adjuster rods irradiated in the CANDU 6 reactor core.

    PubMed

    Gugiu, Daniela; Dumitrache, Ion

    2005-01-01

    The present work is a part of a more complex project related to the replacement of the original stainless steel adjuster rods with cobalt assemblies in the CANDU 6 reactor core. The 60Co produced by 59Co irradiation could be used extensively in medicine and industry. The paper will mainly describe some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronic equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and heating of the irradiated cobalt rods, are performed using the Monte Carlo codes MCNP5 and MONTEBURNS 2.1. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters. PMID:16604599

  1. FAST: A Fuel And Sheath Modeling Tool for CANDU Reactor Fuel

    NASA Astrophysics Data System (ADS)

    Prudil, Andrew Albert

    Understanding the behaviour of nuclear fuel during irradiation is a complicated multiphysics problem involving neutronics, chemistry, radiation physics, material-science, solid mechanics, heat transfer and thermal-hydraulics. Due to the complexity and interdependence of the physics and models involved, fuel modeling is typically clone with numerical models. Advancements in both computer hardware and software have made possible new more complex and sophisticated fuel modeling codes. The Fuel And Sheath modelling Tool (FAST) is a fuel performance code that has been developed for modeling nuclear fuel behaviour under normal and transient conditions. The FAST code includes models for heat generation and transport, thermal expansion, elastic strain, densification, fission product swelling, pellet relocation, contact, grain growth, fission gas release, gas and coolant pressure and sheath creep. These models are coupled and solved numerically using the Comsol Multiphysics finite-element platform. The model utilizes a radialaxial geometry of a fuel pellet (including dishing and chamfering) and accompanying fuel sheath allowing the model to predict circumferential ridging. This model has evolved from previous treatments developed at the Royal Military College. The model has now been significantly advanced to include: a more detailed pellet geometry, localized pellet-to-sheath gap size and contact pressure, ability to model cracked pellets, localized fuel burnup for material property models, improved U02 densification behaviour, fully 2-dimensional model for the sheath, additional creep models, additional material models, an FEM Booth-diffusion model for fission gas release (including ability to model temperature and power changes), a capability for end-of-life predictions, the ability to utilize text files as model inputs, and provides a first time integration of normal operating conditions (NOC) and transient fuel models into a single code (which has never been achieved

  2. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    SciTech Connect

    Forsberg, C.W.; Reich, W.J.

    1991-09-01

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.

  3. Nuclear reactors for space power

    SciTech Connect

    Buden, D.

    1985-02-01

    The growth in power demands for spacecraft, especially outer planet missions, is driving the development of space nuclear power systems. Nuclear reactors could also be used to process lunar materials to take advantage of order of magnitude lower fuel requirements to move construction components off the moon instead of the earth. Larger, more powerful broadcast satellites which lower the GEO station space demand could use nuclear power, as could navigational systems, orbital transfer vehicles and a manned Mars mission. The SP-100 design is currently undergoing parametric evaluation before engineering studies begin. Safety concerns are concentrated on preventing fissioning until the reactor is on-orbit and keeping the active or discarded reactor out of the atmosphere until the radioactivity has decayed to levels defined by international standards.

  4. Multimegawatt space power reactors

    SciTech Connect

    Dearien, J.A.; Whitbeck, J.F.

    1989-01-01

    In response to the need of the Strategic Defense Initiative (SDI) and long range space exploration and extra-terrestrial basing by the National Air and Space Administration (NASA), concepts for nuclear power systems in the multi-megawatt levels are being designed and evaluated. The requirements for these power systems are being driven primarily by the need to minimize weight and maximize safety and reliability. This paper will discuss the present requirements for space based advanced power systems, technological issues associated with the development of these advanced nuclear power systems, and some of the concepts proposed for generating large amounts of power in space. 31 figs.

  5. The U.S.-Russian joint studies on using power reactors to disposition surplus weapon plutonium as spent fuel

    SciTech Connect

    Chebeskov, A.; Kalashnikov, A.; Bevard, B.; Moses, D.; Pavlovichev, A.

    1997-09-01

    In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years.

  6. Neutronic calculations for CANDU thorium systems using Monte Carlo techniques

    NASA Astrophysics Data System (ADS)

    Saldideh, M.; Shayesteh, M.; Eshghi, M.

    2014-08-01

    In this paper, we have investigated the prospects of exploiting the rich world thorium reserves using Canada Deuterium Uranium (CANDU) reactors. The analysis is performed using the Monte Carlo MCNP code in order to understand how much time the reactor is in criticality conduction. Four different fuel compositions have been selected for analysis. We have obtained the infinite multiplication factor, k∞, under full power operation of the reactor over 8 years. The neutronic flux distribution in the full core reactor has already been investigated.

  7. POWER BREEDER REACTOR

    DOEpatents

    Monson, H.O.

    1960-11-22

    An arrangement is offered for preventing or minimizing the contraction due to temperature rise, of a reactor core comprising vertical fuel rods in sodium. Temperature rise of the fuel rods would normally make them move closer together by inward bowing, with a resultant undesired increase in reactivity. According to the present invention, assemblies of the fuel rods are laterally restrained at the lower ends of their lower blanket sections and just above the middle of the fuel sections proper of the rods, and thus the fuel sections move apart, rather than together, with increase in temperature.

  8. Estimation of clearance potential index and hazard factors of Candu fuel bundle and its validation based on the measurements of radioisotopes inventories from Pickering reactor fuel

    SciTech Connect

    Pavelescu, Alexandru Octavian; Tinti, Renato; Voukelatou, Konstantina; Cepraga, Dan Gabriel

    2007-07-01

    This paper is related to the clearance potential levels, ingestion and inhalation hazard factors of the spent nuclear fuel and radioactive wastes. This study required a complex activity that consisted of more steps such as: the acquisition, setting up, validation and application of procedures, codes and libraries. The paper reflects the validation stage of this study. Its objective was to compare the measured inventories of selected actinide and fission products radionuclides in an element from the Pickering Candu reactor with the inventories predicted using a recent version of the SCALE 5/ORIGEN-ARP code coupled with the time dependent cross sections library for the Candu 28 reactor (produced by the sequence SCALE4.4a/SAS2H and SCALE4.4a/ORIGEN-S). In this way, the procedures, the codes and the libraries for the characterization of radioactive material in terns of radioactive inventories, clearance, and biological hazard factors could be qualified and validated, in support of the safety management of the radioactive wastes. (authors)

  9. Significance of analog instrumentation - design philosophy of replacement dump arrest unit at Pickering Station Candu Reactor

    SciTech Connect

    Miller, J.F.; McDowell, R.W.

    1996-12-31

    This paper discusses the differences of opinion concerning power plant instrumentation, including safety systems. One popular view point is that modem instrumentation must be microprocessor-based to be acceptable. An alternative view point is that properly designed analog instrumentation is recommended in some applications and has proven to be viable based upon performance and experience. A practical example is discussed in detail, explaining how a combination of discrete analog circuitry, combined with discrete digital circuitry provides a robust solution to a complex instrumentation replacement problem. In this application, a microprocessor-based instrument was designed as a replacement for an obsolete analog instrument. Due to severe licensing difficulties, the instrument was redesigned as a combination of discrete analog and digital circuitry. In the implementation of this circuitry, all complex testing functions of the improved microprocessor-based instrument were accommodated and system accuracy and performance were not compromised over the micro-processor-based instrument. The instrument has met all requirements for reliability and EMI/RFI susceptibility, as well as isolation of analog outputs and the ability to withstand severe transient noise on inputs and outputs without adversely affecting performance.

  10. PUSH-PULL POWER REACTOR

    DOEpatents

    Froman, D.K.

    1959-02-24

    Power generating nuclear reactors of the homogeneous liquid fuel type are discussed. The apparatus utilizes two identical reactors interconnected by conduits through heat exchanging apparatus. Each reactor contains a critical geometry region and a vapor region separated from the critical region by a baffle. When the liquid in the first critical region becomes critical, the vapor pressure above the fuel is increased due to the rise in the temperature until it forces the liquid fuel out of the first critical region through the heat exchanger and into the second critical region, which is at a lower temperature and consequently a lower vapor pressure. The above reaction is repeated in the second critical region and the liquid fuel is forced back into the first critical region. In this manner criticality is achieved alternately in each critical region and power is extracted by the heat exchanger from the liquid fuel passing therethrough. The vapor region and the heat exchanger have a non-critical geometry and reactivity control is effected by conventional control rods in the critical regions.

  11. Hybrid Reactor Simulation of Boiling Water Reactor Power Oscillations

    SciTech Connect

    Huang Zhengyu; Edwards, Robert M.

    2003-08-15

    Hybrid reactor simulation (HRS) of boiling water reactor (BWR) instabilities, including in-phase and out-of-phase (OOP) oscillations, has been implemented on The Pennsylvania State University TRIGA reactor. The TRIGA reactor's power response is used to simulate reactor neutron dynamics for in-phase oscillation or the fundamental mode of the reactor modal kinetics for OOP oscillations. The reactor power signal drives a real-time boiling channel simulation, and the calculated reactivity feedback is in turn fed into the TRIGA reactor via an experimental changeable reactivity device. The thermal-hydraulic dynamics, together with first harmonic mode power dynamics, is digitally simulated in the real-time environment. The real-time digital simulation of boiling channel thermal hydraulics is performed by solving constitutive equations for different regions in the channel and is realized by a high-performance personal computer. The nonlinearity of the thermal-hydraulic model ensures the capability to simulate the oscillation phenomena, limit cycle and OOP oscillation, in BWR nuclear power plants. By adjusting reactivity feedback gains for both modes, various oscillation combinations can be realized in the experiment. The dynamics of axially lumped power distribution over the core is displayed in three-dimensional graphs. The HRS reactor power response mimics the BWR core-wide power stability phenomena. In the OOP oscillation HRS, the combination of reactor response and the simulated first harmonic power using shaping functions mimics BWR regional power oscillations. With this HRS testbed, a monitoring and/or control system designed for BWR power oscillations can be experimentally tested and verified.

  12. Small reactor power system for space application

    NASA Technical Reports Server (NTRS)

    Shirbacheh, M.

    1987-01-01

    A development history and comparative performance capability evaluation is presented for spacecraft nuclear powerplant Small Reactor Power System alternatives. The choice of power conversion technology depends on the reactor's operating temperature; thermionic, thermoelectric, organic Rankine, and Alkali metal thermoelectric conversion are the primary power conversion subsystem technology alternatives. A tabulation is presented for such spacecraft nuclear reactor test histories as those of SNAP-10A, SP-100, and NERVA.

  13. Power reactor embrittlement data base

    SciTech Connect

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1989-01-01

    Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well-designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: to compile and to verify the quality of the PR-EDB; to provide user-friendly software to access and process the data; to explore or confirm embrittlement prediction models; and to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. 9 figs.

  14. Zirconium Hydride Space Power Reactor design.

    NASA Technical Reports Server (NTRS)

    Asquith, J. G.; Mason, D. G.; Stamp, S.

    1972-01-01

    The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

  15. Thermionic reactors for space nuclear power

    NASA Technical Reports Server (NTRS)

    Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

    1985-01-01

    Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

  16. Experimental development of power reactor advanced controllers

    SciTech Connect

    Edwards, R.M.; Weng, C.K.; Lindsay, R.W.

    1992-06-01

    A systematic approach for developing and verifying advanced controllers with potential application to commercial nuclear power plants is suggested. The central idea is to experimentally demonstrate an advanced control concept first on an ultra safe research reactor followed by demonstration on a passively safe experimental power reactor and then finally adopt the technique for improving safety, performance, reliability and operability at commercial facilities. Prior to completing an experimental sequence, the benefits and utility of candidate advanced controllers should be established through theoretical development and simulation testing. The applicability of a robust optimal observer-based state feedback controller design process for improving reactor temperature response for a TRIGA research reactor, Liquid Metal-cooled Reactor (LMR), and a commercial Pressurized Water Reactor (PWR) is presented to illustrate the potential of the proposed experimental development concept.

  17. Experimental development of power reactor advanced controllers

    SciTech Connect

    Edwards, R.M. . Dept. of Nuclear Engineering); Weng, C.K. . Dept. of Mechanical Engineering); Lindsay, R.W. )

    1992-01-01

    A systematic approach for developing and verifying advanced controllers with potential application to commercial nuclear power plants is suggested. The central idea is to experimentally demonstrate an advanced control concept first on an ultra safe research reactor followed by demonstration on a passively safe experimental power reactor and then finally adopt the technique for improving safety, performance, reliability and operability at commercial facilities. Prior to completing an experimental sequence, the benefits and utility of candidate advanced controllers should be established through theoretical development and simulation testing. The applicability of a robust optimal observer-based state feedback controller design process for improving reactor temperature response for a TRIGA research reactor, Liquid Metal-cooled Reactor (LMR), and a commercial Pressurized Water Reactor (PWR) is presented to illustrate the potential of the proposed experimental development concept.

  18. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    DOEpatents

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  19. Automated refueling simulations of a CANDU for the exploitation of thorium fuels

    NASA Astrophysics Data System (ADS)

    Holmes, Bradford

    CANDU nuclear reactors are in a unique circumstance where they are able to utilize and exploit a number of different fuel options to provide power as a utility. Thorium, a fertile isotope found naturally, is one option that should be explored. Thorium is more abundant than uranium, which is the typical fuel in the reactor and the availability of thorium makes nuclear energy desirable to more countries. This document contains the culmination of a project that explores, tests, and analyzes the feasibility of using thorium in a CANDU reactor. The project first develops a set of twodimensional lattice and three dimensional control rod simulations using the DRAGON Version 4 nuclear physics codes. This step is repeated for many concentrations of thorium. The data generated in these steps is then used to determine a functional enrichment of thorium. This is done via a procedural elimination and optimization of certain key parameters including but not limited to average exit burnup and reactivity evolution. For the purposes of this project, an enrichment of 1 % thorium was found viable. Full core calculations were done using the DONJON 4 code. CANFUEL, a program which simulates the refueling operations of a CANDU reactor for this fuel type was developed and ran for a simulation period of one hundred days. The program and the fuel selection met all selected requirements for the entirety of the simulation period. CANFUEL requires optimization for fuel selection before it can be used extensively. The fuel selection was further scrutinized when a reactivity insertion event was simulated. The adjuster rod 11 withdrawal from the core was analyzed and compared to classical CANDU results in order to ensure no significant deviations or unwanted evolutions were encountered. For this case, the simulation results were deemed acceptable with no significant deviations from the classical CANDU case.

  20. Low power reactor for remote applications

    SciTech Connect

    Meier, K.L.; Palmer, R.G.; Kirchner, W.L.

    1985-01-01

    A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long-term, virtually maintenance free, operation of this reactor for remote applications. 10 refs., 7 figs., 3 tabs.

  1. Low power reactor for remote applications

    NASA Astrophysics Data System (ADS)

    Meier, K. L.; Palmer, R. G.; Kirchner, W. L.

    1985-05-01

    A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long term, virtually maintenance free, operation of this reactor for remote applications.

  2. Reactor power system/spacecraft integration

    NASA Technical Reports Server (NTRS)

    Elms, R. V.

    1985-01-01

    The new national initiative in space reactor technology evaluation and development is strongly tied to mission applications and to spacecraft and space transportation system (STS) compatibility. This paper discusses the power system integration interfaces with potential using spacecraft and the STS, and the impact of these requirements on the design. The integration areas of interest are mechanical, thermal, electrical, attitude control, and mission environments. The mission environments include space vacuum, solar input, heat sink, space radiation, weapons effects, and reactor power system radiation environments. The natural, reactor, and weapons effects radiation must be evaluated and combined to define the design requirements for spacecraft electronic equipment.

  3. Compact reactor/ORC power source

    SciTech Connect

    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/sup 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/sup 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.

  4. Heat pipe reactors for space power applications

    NASA Technical Reports Server (NTRS)

    Koenig, D. R.; Ranken, W. A.; Salmi, E. W.

    1977-01-01

    A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kWe and operate in the temperature range 1200-1700 K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO2. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor. Virtues of the reactor designs are the avoidance of single-point failure mechanisms, the relatively high operating temperature, and the expected long lifetimes of the fuel element components.

  5. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  6. POWER GENERATING NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Vernon, H.C.

    1958-03-01

    This patent relates to reactor systems of the type wherein the cooiing medium is a liquid which is converted by the heat of the reaction to steam which is conveyed directly to a pnime mover such as a steam turbine driving a generatore after which it is condensed and returred to the coolant circuit. In this design, the reactor core is disposed within a tank for containing either a slurry type fuel or an aggregation of solid fuel elements such as elongated rods submerged in a liquid moderator such as heavy water. The top of the tank is provided with a nozzle which extends into an expansion chamber connected with the upper end of the tank, the coolant being maintained in the expansion chamber at a level above the nozzle and the steam being formed in the expansion chamber.

  7. Advances in Tandem Mirror fusion power reactors

    SciTech Connect

    Perkins, L.J.; Logan, B.G.

    1986-05-20

    The Tandem Mirror exhibits several distinctive features which make the reactor embodiment of the principle very attractive: Simple low-technology linear central cell; steady-state operation; high-..beta.. operation; no driven current or disruptions; divertorless operation; direction conversion of end-loss power; low-surface heat loads; and advanced fusion fuel capability. In this paper, we examine these features in connection with two tandem mirror reactor designs, MARS and MINIMARS, and several advanced reactor concepts including the wall-stabilized reactor and the field-reversed mirror. With a novel compact end plug scheme employing octopole stabilization, MINIMARS is expressly designed for short construction times, factory-built modules, and a small (600 MWe) but economic reactor size. We have also configured the design for low radioactive afterheat and inherent/passive safety under LOCA/LOFA conditions, thereby obviating the need for expensive engineered safety systems. In contrast to the complex and expensive double-quadrupole end-cell of the MARS reactor, the compact octopole end-cell of MINIMARS enables ignition to be achieved with much shorter central cell lengths and considerably improves the economy of scale for small (approx.250 to 600 MWe) tandem mirror reactors. Finally, we examine the prospects for realizing the ultimate potential of the tandem mirror with regard to both innovative configurations and novel neutron energy conversion schemes, and stress that advanced fuel applications could exploit its unique reactor features.

  8. Liquid Metal Cooled Reactor for Space Power

    NASA Astrophysics Data System (ADS)

    Weitzberg, Abraham

    2003-01-01

    The conceptual design is for a liquid metal (LM) cooled nuclear reactor that would provide heat to a closed Brayton cycle (CBC) power conversion subsystem to provide electricity for electric propulsion thrusters and spacecraft power. The baseline power level is 100 kWe to the user. For long term power generation, UN pin fuel with Nb1Zr alloy cladding was selected. As part of the SP-100 Program this fuel demonstrated lifetime with greater than six atom percent burnup, at temperatures in the range of 1400-1500 K. The CBC subsystem was selected because of the performance and lifetime database from commercial and aircraft applications and from prior NASA and DOE space programs. The high efficiency of the CBC also allows the reactor to operate at relatively low power levels over its 15-year life, minimizing the long-term power density and temperature of the fuel. The scope of this paper is limited to only the nuclear components that provide heated helium-xenon gas to the CBC subsystem. The principal challenge for the LM reactor concept was to design the reactor core, shield and primary heat transport subsystems to meet mission requirements in a low mass configuration. The LM concept design approach was to assemble components from prior programs and, with minimum change, determine if the system met the objective of the study. All of the components are based on technologies having substantial data bases. Nuclear, thermalhydraulic, stress, and shielding analyses were performed using available computer codes. Neutronics issues included maintaining adequate operating and shutdown reactivities, even under accident conditions. Thermalhydraulic and stress analyses calculated fuel and material temperatures, coolant flows and temperatures, and thermal stresses in the fuel pins, components and structures. Using conservative design assumptions and practices, consistent with the detailed design work performed during the SP-100 Program, the mass of the reactor, shield, primary heat

  9. Gas-core reactor power transient analysis.

    NASA Technical Reports Server (NTRS)

    Kascak, A. F.

    1972-01-01

    The nuclear fuel in the gas-core reactor concept is a ball of uranium plasma radiating thermal photons. The photons are met by an inflowing hydrogen stream, which is seeded with submicron size, depleted uranium particles. A 'wall-burnout' condition exists if the thermal photons can reach the cavity liner because of insufficient absorption by the hydrogen. An analysis was conducted in order to determine the time for which the maximum steady state reactor power could be exceeded without damage to the cavity liner due to burnout. Wall-burnout time as a function of the power increase above the initial steady state condition is shown in a graph.

  10. Reactor power system deployment and startup

    NASA Technical Reports Server (NTRS)

    Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

    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.

  11. A small, 1400 K, reactor for Brayton space power systems.

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Mayo, W.

    1972-01-01

    An investigation was conducted to determine minimum dimensions and minimum weight obtainable in a design for a reactor using uranium-233 nitride or plutonium-239 nitride as fuel. Such a reactor had been considered by Krasner et al. (1971). Present space power status is discussed, together with questions of reactor design and power distribution in the reactor. The characteristics of various reactor types are compared, giving attention also to a zirconium hydride reactor.

  12. Korean experience in CANDU-PHWR operation

    SciTech Connect

    Park, S.K.

    1988-01-01

    Among KEPCO's 9 nuclear power units, Korea Nuclear Unit No. 3, the Wolsung Nuclear Power Plant is the only CANDU-PHWR Unit, while the rest of 8 others are PWR units. The unit was designed by Atomic Energy of Canada, Ltd. of Canada, who also performed overall project management for the plant construction under the provisions and arrangement of the relevant contracts. The gross electrical output of the plant is 678.7 MWe and thermal output of the reactor is 2061 MWth. While these figures lead to lower plant efficiency than LWR counterparts, unit energy cost for fuel is more favorable than LWRs because natural uranium is utilized for the fuel bundles, some of which are already being fabricated domestically. Annual capacity factors for 1983 and 1984 could have been improved, if two major planned outages for the modification works on steam generator internals and one major forced outage from the heavy water spill incident could be eliminated. The heavy water spill incident in November, 1984 brought plant staffs many lessons to learn and many things to contemplate. Unique design concepts and features such as on-power refuelling, poison prevent mode, versatile plant control system built around digital computers and power step back/set back logics may be credited for these relatively good plant performances. Human related factors such as staff's technical capabilities and strong will toward good performance were other elements which could not be overlooked.

  13. Static conversion systems. [for space power reactors

    NASA Technical Reports Server (NTRS)

    Ewell, R.; Mondt, J.

    1985-01-01

    Historically, all space power systems that have actually flown in space have relied on static energy conversion technology. Thus, static conversion is being considered for space nuclear power systems as well. There are four potential static conversion technologies which should be considered. These include: the alkali metal thermoelectric converter (AMTEC), the thermionic converter, the thermoelectric converter, and the thermophotovoltaic converter (TPV). These four conversion technologies will be described in brief detail along with their current status and development needs. In addition, the systems implications of using each of these conversion technologies with a space nuclear reactor power system will be evaluated and some comparisons made.

  14. A prototype expert system for the monitoring of defected nuclear fuel elements in Canada deuterium uranium reactors

    SciTech Connect

    Lewis, B.J.; Green, R.J. ); Che, C.W.T. )

    1992-06-01

    This paper reports on a prototype expert system for fuel failure monitoring in Canada deuterium uranium (CANDU) power reactors. Based on a coolant activity analysis, the system is able to provide information in an operating reactor on the number of fuel failures, the average defect size, and the amount of tramp uranium deposited on the in-core surfaces of the primary heat transport system. The fission product release model used in the system is based on results from an in-reactor experimental program at Chalk River Nuclear Laboratories. The expert system is validated against fuel failure data from a number of CANDU power reactors.

  15. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    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.

  16. Gas-core reactor power transient analysis

    NASA Technical Reports Server (NTRS)

    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 this 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.

  17. Neutronics-thermalhydraulics coupling in a CANDU SCWR

    NASA Astrophysics Data System (ADS)

    Adouki, Pierre

    In order to implement new nuclear technologies as a solution to the growing demand for energy, 10 countries agreed on a framework for international cooperation in 2002, to form the Generation IV International Forum (GIF). The goal of the GIF is to design the next generation of nuclear reactors that would be cost effective and would enhance safety. This forum has proposed several types of Generation IV reactors including the Supercritical Water-Cooled Reactor (SCWR). The SCWR comes in two main configurations: pressure vessel SCWR and pressure tube SCWR (PT-SCWR). In this study, the CANDU SCWR (a PT-SCWR) is considered. This reactor is oriented vertically and contains 336 channels with a length of 5 m. The target coolant inlet and outlet temperatures are 350 Celsius and 625 Celsius, respectively. The coolant flows downwards, and the reactor power is 2540 MWth. Various fuel designs have been considered in order not to exceed the linear element rating. However, the dependency between the core power and thermalhydraulics parameters results in the necessity to use a neutronics/thermalhydaulics coupling scheme to determine the core power and the thermalhydraulics parameters. The core power obtained has a power peaking factor of 1.4. The bundle power distribution for all channels has a peak at the third bundle from the inlet, but the value of this peak increases with the channel power. The heat-transfer coefficient and the specific-heat capacity have a peak at the same location in a channel, and this location shifts toward the inlet as the channel power increases. The exit coolant temperature increases with the channel power, while the exit coolant density and pressure decrease with the channel power. Also, higher channel powers lead to higher fuel and cladding temperatures. Moreover, as the coupling method is applied, the effective multiplication factor and the values of thermalhydaulics parameters oscillate as they converge.

  18. NACA Zero Power Reactor Facility Hazards Summary

    NASA Technical Reports Server (NTRS)

    1957-01-01

    The Lewis Flight Propulsion Laboratory of the National Advisory Committee for Aeronautics proposes to build a zero power research reactor facility which will be located in the laboratory grounds near Clevelaurd, Ohio. The purpose of this report is to inform the Advisory Commit tee on Reactor Safeguards of the U. S. Atomic Energy Commission in re gard to the design of the reactor facility, the cha,acteristics of th e site, and the hazards of operation at this location, The purpose o f this reactor is to perform critical experiments, to measure reactiv ity effects, to serve as a neutron source, and to serve as a training tool. The reactor facility is described. This is followed by a discu ssion of the nuclear characteristics and the control system. Site cha racteristics are then discussed followed by a discussion of the exper iments which may be conducted in the facility. The potential hazards of the facility are then considered, particularly, the maximum credib le accident. Finally, the administrative procedure is discussed.

  19. 77 FR 38742 - Non-Power Reactor License Renewal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-29

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN 3150-AI96 Non-Power Reactor License Renewal AGENCY... renewal requirements for non-power reactors. This contemplated rulemaking would also make conforming changes to address technical issues in existing non-power reactor regulations. The NRC is seeking...

  20. BN-800 advanced nuclear power plant with fast reactor

    SciTech Connect

    Shishkin, A.N.; Kuzavkov, N.G.; Sobolev, V.A.; Shestakov, G.V.; Bagdasarov, Yu.E.; Kochetkov, L.A.; Matveyev, V.I.; Poplavsky, V.M.

    1993-12-31

    Bn-800 reactor plant with fast reactor and sodium coolant in the primary and secondary circuits is designed for operation as part of the power units in the Yuzhno-Uralskaya nuclear power plant scheduled to be constructed in Chelyabinsk region and as part unit 4 in the Beloyarskaya nuclear power plant. Reactor operations are described.

  1. Enhanced Singular Wave Reactor for Surface Power

    NASA Astrophysics Data System (ADS)

    Popa-Simil, L.

    The "CANDLE" (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) also known as singular wave reactor has many significant advantages related to elimination of the need for enrichment. The use of micro-hetero structured fuel, generically called "cer-liq-mesh" will further improve burnup up to 90%. In spite it has typically large dimensions, being heavy to be transported in space, in a single piece, but because it will deliver energy in hundreds MW level for about 100 years per charge using natural Uranium or Thorium as fuel available on the planet's surface, and because it can be assembled locally becomes a very attractive option for self sustainable power cycles. The "cer-liq-mesh" fuel based singular wave reactor is smaller, less than ¼ from the size of "Candle" reactor, and has a very high burnup reducing the fuel cycle drastically. It can be transported by parts, with extremely small probability of over-unity criticality accident and be assembled to run on the surface. This represents a better option for extraterrestrial applications; in spite it requires a more complicated fuel fabrication that pays back in a simplified fuel cycle and minimum waste.

  2. Modular stellarator reactor: a fusion power plant

    SciTech Connect

    Miller, R.L.; Bathke, C.G.; Krakowski, R.A.; Heck, F.M.; Green, L.; Karbowski, J.S.; Murphy, J.H.; Tupper, R.B.; DeLuca, R.A.; Moazed, A.

    1983-07-01

    A comparative analysis of the modular stellarator and the torsatron concepts is made based upon a steady-state ignited, DT-fueled, reactor embodiment of each concept for use as a central electric-power station. Parametric tradeoff calculations lead to the selection of four design points for an approx. 4-GWt plant based upon Alcator transport scaling in l = 2 systems of moderate aspect ratio. The four design points represent high-aspect ratio. The four design points represent high-(0.08) and low-(0.04) beta versions of the modular stellarator and torsatron concepts. The physics basis of each design point is described together with supporting engineering and economic analyses. The primary intent of this study is the elucidation of key physics and engineering tradeoffs, constraints, and uncertainties with respect to the ultimate power reactor embodiment.

  3. Power conditioning for space nuclear reactor systems

    NASA Technical Reports Server (NTRS)

    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.

  4. Assessment for advanced fuel cycle options in CANDU

    SciTech Connect

    Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

    2013-07-01

    The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a driver fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.

  5. Nuclear safety as applied to space power reactor systems

    SciTech Connect

    Cummings, G.E.

    1987-01-01

    Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety.

  6. Analysis of UF6 breeder reactor power plants

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.

    1976-01-01

    Gaseous UF6 fueled breeder reactor design and technical applications of such concepts are summarized. Special attention was given to application in nuclear power plants and to reactor efficiency and safety factors.

  7. REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

    DOEpatents

    Loeb, E.

    1961-01-17

    A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

  8. Nuclear reactor power for an electrically powered orbital transfer vehicle

    NASA Technical Reports Server (NTRS)

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

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low earth orbit (LEO) and geosynchronous earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant.

  9. UF6 breeder reactor power plants for electric power generation

    NASA Technical Reports Server (NTRS)

    Rust, J. H.; Clement, J. D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

  10. Thermionic reactor power conditioner design for nuclear electric propulsion.

    NASA Technical Reports Server (NTRS)

    Jacobsen, A. S.; Tasca, D. M.

    1971-01-01

    Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

  11. 78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-09

    ... COMMISSION Operator Licensing Examination Standards for Power Reactors AGENCY: Nuclear Regulatory Commission... Standards for Power Reactors.'' DATES: Submit comments by February 7, 2014. Comments received after this..., and grading of examinations used for licensing operators at nuclear power plants pursuant to...

  12. Tokamak power reactor ignition and time dependent fractional power operation

    SciTech Connect

    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.

  13. SP-100 Space Reactor Power System readiness

    NASA Astrophysics Data System (ADS)

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

    The SP-100 Space Reactor Power System 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). An effective infrastructure of Industry, National Laboratories and Government agencies has made substantial progress since the 1988 System Design Review. Hardware development and testing has progressed to the point of resolving all key technical feasibility issues. 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.

  14. Programmable AC power supply for simulating power transient expected in fusion reactor

    SciTech Connect

    Halimi, B.; Suh, K. Y.

    2012-07-01

    This paper focus on control engineering of the programmable AC power source which has capability to simulate power transient expected in fusion reactor. To generate the programmable power source, AC-AC power electronics converter is adopted to control the power of a set of heaters to represent the transient phenomena of heat exchangers or heat sources of a fusion reactor. The International Thermonuclear Experimental Reactor (ITER) plasma operation scenario is used as the basic reference for producing this transient power source. (authors)

  15. Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs)

    NASA Astrophysics Data System (ADS)

    Lizon-A-Lugrin, Laure

    The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a

  16. Assessment of nuclear reactor concepts for low power space applications

    NASA Technical Reports Server (NTRS)

    Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

    1988-01-01

    The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

  17. 77 FR 60039 - Non-Power Reactor License Renewal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-02

    ... FR 38742), for comment from the public, licensees, certificate holders, and other stakeholders. The... COMMISSION 10 CFR Part 50 RIN 3150-AI96 Non-Power Reactor License Renewal AGENCY: Nuclear Regulatory... streamline non-power reactor license renewal. This final regulatory basis incorporates input from the...

  18. A method of examining iron oxides speciation and transport to steam generators during nuclear power reactor startups

    NASA Astrophysics Data System (ADS)

    Sawicki, Jerzy A.; Sawicka, Barbara D.; Price, James E.

    2010-12-01

    Secondary side corrosion products (sludge) collected during one of CANDU1 reactor startups from wet layup have been examined by X-ray fluorescence and Mössbauer spectroscopy. The transport and chemical form of iron oxides and oxyhydroxides were determined in condensate, feedwater and preheater outlet as a function of temperature and time. The sludge burst and oxidation states of iron oxides were correlated with the rise of reactor power and corresponding changes in temperature, condensate vacuum and water flow rate. In particular, a sharp γ-FeOOH to Fe 3O 4 switch was observed that coincided in time with the onset of condensate vacuum. Also, it was found that the startup after wet layup is characterized by only brief and fairly small sludge burst at about 30% reactor power and which contributes only a small amount of undesirable α-Fe 2O 3 to total iron transport to steam generator. Thus, sludge burden to steam generators can be minimized with proper layup and startup practices. ™ Trademark of Atomic Energy of Canada Limited.

  19. Power ascension strategy following a reactor trip during EOC coastdown

    SciTech Connect

    Beard, C.L.; Heibel, M.D. ); Lesnick, D.C. )

    1992-01-01

    The difficulties associated with returning a reactor to the pretrip power level following a reactor trip during an end-of-cycle (EOC) power coastdown maneuver, and maintaining it once achieved, have caused utilities to abandon the restart and enter their refueling outages ahead of schedule. The Commonwealth Edison Company (CECo) Braidwood and Byron units have experienced reactor trips during EOC power coastdown maneuvers and have successfully performed restarts. The installation of the BEACON core monitoring system, which provides core monitoring, measurement reduction, core analysis and follow, and core prediction capability utilizing a very fast and accurate three-dimensional nodal code, at the CECo Byron, Braidwood, and Zion stations allows the reactor engineers at these units to accurately determine reactor response. The capabilities of the BEACON system allow an optimal return to power strategy to be developed and continuously updated. This paper presents a method for establishing the optimal return to power strategy utilizing the BEACON system.

  20. Axial power monitoring uncertainty in the Savannah River Reactors

    SciTech Connect

    Losey, D.C.; Revolinski, S.M.

    1990-01-01

    The results of this analysis quantified the uncertainty associated with monitoring the Axial Power Shape (APS) in the Savannah River Reactors. Thermocouples at each assembly flow exit map the radial power distribution and are the primary means of monitoring power in these reactors. The remaining uncertainty in power monitoring is associated with the relative axial power distribution. The APS is monitored by seven sensors that respond to power on each of nine vertical Axial Power Monitor (APM) rods. Computation of the APS uncertainty, for the reactor power limits analysis, started with a large database of APM rod measurements spanning several years of reactor operation. A computer algorithm was used to randomly select a sample of APSs which were input to a code. This code modeled the thermal-hydraulic performance of a single fuel assembly during a design basis Loss-of Coolant Accident. The assembly power limit at Onset of Significant Voiding was computed for each APS. The output was a distribution of expected assembly power limits that was adjusted to account for the biases caused by instrumentation error and by measuring 7 points rather than a continuous APS. Statistical analysis of the final assembly power limit distribution showed that reducing reactor power by approximately 3% was sufficient to account for APS variation. This data confirmed expectations that the assembly exit thermocouples provide all information needed for monitoring core power. The computational analysis results also quantified the contribution to power limits of the various uncertainties such as instrumentation error.

  1. Axial power monitoring uncertainty in the Savannah River Reactors

    SciTech Connect

    Losey, D.C.; Revolinski, S.M.

    1990-12-31

    The results of this analysis quantified the uncertainty associated with monitoring the Axial Power Shape (APS) in the Savannah River Reactors. Thermocouples at each assembly flow exit map the radial power distribution and are the primary means of monitoring power in these reactors. The remaining uncertainty in power monitoring is associated with the relative axial power distribution. The APS is monitored by seven sensors that respond to power on each of nine vertical Axial Power Monitor (APM) rods. Computation of the APS uncertainty, for the reactor power limits analysis, started with a large database of APM rod measurements spanning several years of reactor operation. A computer algorithm was used to randomly select a sample of APSs which were input to a code. This code modeled the thermal-hydraulic performance of a single fuel assembly during a design basis Loss-of Coolant Accident. The assembly power limit at Onset of Significant Voiding was computed for each APS. The output was a distribution of expected assembly power limits that was adjusted to account for the biases caused by instrumentation error and by measuring 7 points rather than a continuous APS. Statistical analysis of the final assembly power limit distribution showed that reducing reactor power by approximately 3% was sufficient to account for APS variation. This data confirmed expectations that the assembly exit thermocouples provide all information needed for monitoring core power. The computational analysis results also quantified the contribution to power limits of the various uncertainties such as instrumentation error.

  2. Diversion assumptions for high-powered research reactors

    SciTech Connect

    Binford, F.T.

    1984-01-01

    This study deals with diversion assumptions for high-powered research reactors -- specifically, MTR fuel; pool- or tank-type research reactors with light-water moderator; and water, beryllium, or graphite reflectors, and which have a power level of 25 MW(t) or more. The objective is to provide assistance to the IAEA in documentation of criteria and inspection observables related to undeclared plutonium production in the reactors described above, including: criteria for undeclared plutonium production, necessary design information for implementation of these criteria, verification guidelines including neutron physics and heat transfer, and safeguards measures to facilitate the detection of undeclared plutonium production at large research reactors.

  3. Consumption of the electric power inside silent discharge reactors

    SciTech Connect

    Yehia, Ashraf

    2015-01-15

    An experimental study was made in this paper to investigate the relation between the places of the dielectric barriers, which cover the surfaces of the electrodes in the coaxial cylindrical reactors, and the rate of change of the electric power that is consumed in forming silent discharges. Therefore, silent discharges have been formed inside three coaxial cylindrical reactors. The dielectric barriers in these reactors were pasted on both the internal surface of the outer electrode in the first reactor and the external surface of the inner electrode in the second reactor as well as the surfaces of the two electrodes in the third reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at normal temperature and pressure, in parallel with the application of a sinusoidal ac voltage between the electrodes of the reactor. The electric power consumed in forming the silent discharges inside the three reactors was measured as a function of the ac peak voltage. The validity of the experimental results was investigated by applying Manley's equation on the same discharge conditions. The results have shown that the rate of consumption of the electric power relative to the ac peak voltage per unit width of the discharge gap improves by a ratio of either 26.8% or 80% or 128% depending on the places of the dielectric barriers that cover the surfaces of the electrodes inside the three reactors.

  4. The next generation of power reactors - safety characteristics

    SciTech Connect

    Modro, S.M.

    1995-01-01

    The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs.

  5. Long Term Storage with Surveillance of Canadian Prototype Nuclear Power Reactors

    SciTech Connect

    Janzen, Rick

    2008-01-15

    Atomic Energy of Canada (AECL) was originally formed by the government of Canada in 1952 to perform research in radiation and nuclear areas. In the mid 1950's Canada decided to limit itself to peaceful uses of nuclear energy and AECL embarked on several research and development programs, one of them being the development of nuclear power plants. This led to the development of the CANDU{sup TM} design of heavy water power reactors, of which there are now 29 operating around the world. This presentation discusses the present state of the first two CANDU{sup TM} prototype reactors and a prototype boiling light water reactor and lessons learnt after being in a long-term storage with surveillance state for more than 20 years. AECL facilities undergo decommissioning by either a prompt or a deferred removal approach. Both approaches are initiated after an operating facility has been declared redundant and gone through final operational shutdown. For the deferred approach, initial decommissioning activities are performed to put the facility into a sustainable, safe, shutdown state to minimize the hazards and costs of the ensuing extended storage with surveillance (SWS) or Safestor phase. At the appropriate time, the facility is dismantled and removed, or put into a suitable condition for re-use. AECL has a number of facilities that were built during its history, and some of these are now redundant or will become redundant in the near future. The deferred removal approach is part of AECL's decommissioning strategy for several reasons: 1. Reduction in radiation doses to workers during the final dismantling, 2. No facilities are available yet in Canada for the management of quantity of wastes arising from decommissioning, 3. Financial constraints presented by the number of facilities that will undergo decommissioning, compared to the availability of funds to carry out the work. This has led to the development of a comprehensive decommissioning plan that includes all of AECL

  6. Status of power-reactor projects undergoing licensing review

    SciTech Connect

    Silver, E.G.

    1982-07-01

    Recent regulatory and other actions relating to power reactors undergoing licensing review are summarized in Table 1 as of May 1, 1982. Except as otherwise noted, all the information presented in this article is taken from NRC press releases or from the reactor docket file, both of which are available at the NRC Public Document Room, 1717 H Street NW, Washington, DC 20555.

  7. Ultrasonic level and temperature sensor for power reactor applications

    SciTech Connect

    Dress, W.B.: Miller, G.N.

    1983-01-01

    An ultrasonic waveguide employing torsional and extensional acoustic waves has been developed for use as a level and temperature sensor in pressurized and boiling water nuclear power reactors. Features of the device include continuous measurement of level, density, and temperature producing a real-time profile of these parameters along a chosen path through the reactor vessel.

  8. Automatic control system by power distribution in a power-generating reactor

    SciTech Connect

    Aleksakov, A.N.; Podlazov, L.N.; Ryabov, V.I.; Shevchenko, V.V.; Postnikov, V.V.

    1980-12-01

    The development of the theoretical principles of construction of these systems and of sufficiently detailed nonlinear dynamic numerical models of a power-generation unit with an RBMK reactor have allowed a consistent procedure to be produced for the engineering synthesis of an (local automated control) LAC-LEP (local emergency protection) system. The LAC system facilitates the shaping and maintenance of the desired power distribution in the whole volume of the reactor. In emergency situations, the LAC-LEP system qualitatively reduces the power to a safe level and effectively suppresses the power warpings in one-half of the reactor, which are characteristic for these reactors.

  9. Computer optimization of reactor-thermoelectric space power systems

    NASA Technical Reports Server (NTRS)

    Maag, W. L.; Finnegan, P. M.; Fishbach, L. H.

    1973-01-01

    A computer simulation and optimization code that has been developed for nuclear space power systems is described. The results of using this code to analyze two reactor-thermoelectric systems are presented.

  10. Heat pipe nuclear reactor for space power

    NASA Technical Reports Server (NTRS)

    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.

  11. COUPLED FAST-THERMAL POWER BREEDER REACTOR

    DOEpatents

    Avery, R.

    1961-07-18

    A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

  12. Reference Reactor Module for the Affordable Fission Surface Power System

    SciTech Connect

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

    2008-01-21

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). 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. This paper describes the reference AFSPS 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, liquid metal-cooled 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. One of the important 'affordability' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  13. Gas-cooled reactor power systems for space

    SciTech Connect

    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.

  14. Small reactor power systems for manned planetary surface bases

    NASA Astrophysics Data System (ADS)

    Bloomfield, Harvey 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.

  15. Small reactor power systems for manned planetary surface bases

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

    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.

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

    SciTech Connect

    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{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.

  17. Design of megawatt power level heat pipe reactors

    SciTech Connect

    Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao; Reid, Robert Stowers

    2015-11-12

    An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors. The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.

  18. Application of reactor-pumped lasers to power beaming

    SciTech Connect

    Repetti, T.E.

    1991-10-01

    Power beaming is the concept of centralized power generation and distribution to remote users via energy beams such as microwaves or laser beams. The power beaming community is presently performing technical evaluations of available lasers as part of the design process for developing terrestrial and space-based power beaming systems. This report describes the suitability of employing a nuclear reactor-pumped laser in a power beaming system. Although there are several technical issues to be resolved, the power beaming community currently believes that the AlGaAs solid-state laser is the primary candidate for power beaming because that laser meets the many design criteria for such a system and integrates well with the GaAs photodiode receiver array. After reviewing the history and physics of reactor-pumped lasers, the advantages of these lasers for power beaming are discussed, along with several technical issues which are currently facing reactor-pumped laser research. The overriding conclusion is that reactor-pumped laser technology is not presently developed to the point of being technially or economically competitive with more mature solid-state technologies for application to power beaming. 58 refs.

  19. Application of reactor-pumped lasers to power beaming

    NASA Astrophysics Data System (ADS)

    Repetti, T. E.

    1991-10-01

    Power beaming is the concept of centralized power generation and distribution to remote users via energy beams such as microwaves or laser beams. The power beaming community is presently performing technical evaluations of available lasers as part of the design process for developing terrestrial and space-based power beaming systems. This report describes the suitability of employing a nuclear reactor-pumped laser in a power beaming system. Although there are several technical issues to be resolved, the power beaming community currently believes that the AlGaAs solid-state laser is the primary candidate for power beaming because that laser meets the many design criteria for such a system and integrates well with the GaAs photodiode receiver array. After reviewing the history and physics of reactor-pumped lasers, the advantages of these lasers for power beaming are discussed, along with several technical issues which are currently facing reactor-pumped laser research. The overriding conclusion is that reactor-pumped laser technology is not presently developed to the point of being technically or economically competitive with more mature solid-state technologies for application to power beaming.

  20. A high-fidelity Monte Carlo evaluation of CANDU-6 safety parameters

    SciTech Connect

    Kim, Y.; Hartanto, D.

    2012-07-01

    Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANDU-6 (CANada Deuterium Uranium) reactor have been evaluated by using a modified MCNPX code. For accurate analysis of the parameters, the DBRC (Doppler Broadening Rejection Correction) scheme was implemented in MCNPX in order to account for the thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted by using the MCNPX and the FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated by using several cross section libraries such as ENDF/B-VI, ENDF/B-VII, JEFF, JENDLE. The PCR value is also evaluated at mid-burnup conditions to characterize safety features of equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, huge number of neutron histories are considered in this work and the standard deviation of the k-inf values is only 0.5{approx}1 pcm. It has been found that the FTC is significantly enhanced by accounting for the Doppler broadening of scattering resonance and the PCR are clearly improved. (authors)

  1. Gas-cooled reactor for space power systems

    SciTech Connect

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors.

  2. Nuclear Power from Fission Reactors. An Introduction.

    ERIC Educational Resources Information Center

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  3. Measurements of the reactor neutron power in absolute units

    NASA Astrophysics Data System (ADS)

    Lebedev, G. V.

    2015-12-01

    The neutron power of the reactor of the Yenisei space nuclear power plant is measured in absolute units using the modernized method of correlation analysis during the ground-based tests of the Yenisei prototypes. Results of the experiments are given. The desired result is obtained in a series of experiments carried out at the stage of the plant preparation for tests. The acceptability of experimental data is confirmed by the results of measuring the reactor neutron power in absolute units at the nominal level by the thermal balance during the life cycle tests of the ground prototypes.

  4. Measurements of the reactor neutron power in absolute units

    SciTech Connect

    Lebedev, G. V.

    2015-12-15

    The neutron power of the reactor of the Yenisei space nuclear power plant is measured in absolute units using the modernized method of correlation analysis during the ground-based tests of the Yenisei prototypes. Results of the experiments are given. The desired result is obtained in a series of experiments carried out at the stage of the plant preparation for tests. The acceptability of experimental data is confirmed by the results of measuring the reactor neutron power in absolute units at the nominal level by the thermal balance during the life cycle tests of the ground prototypes.

  5. High power density reactors based on direct cooled particle beds

    SciTech Connect

    Powell, J.R.; Horn, F.L.

    1985-01-01

    Reactors based on direct cooled HTGR type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out long the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBR's) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed. 12 figs.

  6. Protective actions as a factor in power reactor siting

    SciTech Connect

    Gant, K.S.; Schweitzer, M.

    1984-06-01

    This report examines the relationship between a power reactor site and the ease of implementing protective actions (emergency measures a serious accident). Limiting populating density around a reactor lowers the number of people at risk but cannot assure that all protective actions are possible for those who reside near the reactor. While some protective measures can always be taken (i.e., expedient respiratory protection, sheltering) the ability to evacuate the area or find adequate shelter may depend on the characteristics of the area near the reactor site. Generic siting restrictions designed to identify and eliminate these site-specific constraints would be difficult to formulate. The authors suggest identifying possible impediments to protective actions at a proposed reactor site and addressing these problems in the emergency plans. 66 references, 6 figures, 8 tables.

  7. Reactor Lithium Heat Pipes for HP-STMCs Space Reactor Power System

    NASA Astrophysics Data System (ADS)

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2004-02-01

    Design and performance analysis of the nuclear reactor's lithium heat pipes for a 110-kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The evaporator length of the heat pipes is the same as the active core height (0.45 m) and the C-C finned condenser is of the same length as the STMC panels (1.5 m). The C-C finned condenser section is radiatively coupled to the collector shoes of the STMCs placed on both sides. The lengths of the adiabatic section, the values of the power throughput and the evaporator wall temperature depend on the radial location of the heat pipe in the reactor core and the number and dimensions of the potassium heat pipes in the heat rejection radiator. The reactor heat pipes have a total length that varies from 7.57 to 7.73 m, and a 0.2 mm thick Mo-14%Re wick with an average pore radius of 12 μm. The wick is separated from the Mo-14%Re wall by a 0.5 mm annulus filled with liquid lithium, to raise the prevailing capillary limit. The nominal evaporator (or reactor) temperature varies from 1513 to 1591 K and the thermal power of the reactor is 1.6 MW, which averages 12.7 kW for each of the 126 reactor heat pipes. The power throughput per heat pipe increase to a nominal 15.24 kW at the location of the peak power in the core and to 20.31 kW when an adjacent heat pipe fails. The prevailing capillary limit of the reactor heat pipes is 28.3 kW, providing a design margin >= 28%.

  8. New reactor technology: safety improvements in nuclear power systems.

    PubMed

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems. PMID:18049233

  9. Technological Transfer from Research Nuclear Reactors to New Generation Nuclear Power Reactors

    SciTech Connect

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-21

    The goal of this paper is the analysis of the technological transfer role in the nuclear field, with particular emphasis on nuclear reactors domain. The presentation is sustained by historical arguments. In this frame, it is very important to start with the achievements of the first nuclear systems, for instant those with natural uranium as fuel and heavy water as moderator, following in time through the history until the New Generation Nuclear Power Reactors.Starting with 1940, the accelerated development of the industry has implied the increase of the global demand for energy. In this respect, the nuclear energy could play an important role, being essentially an unlimited source of energy. However, the nuclear option faces the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide, a significant amount of experience has been accumulated during development, licensing, construction, and operation of nuclear power reactors. The experience gained is a strong basis for further improvements. Actually, the nuclear programs of many countries are addressing the development of advanced reactors, which are intended to have better economics, higher reliability, improved safety, and proliferation-resistant characteristics in order to overcome the current concerns about nuclear power. Advanced reactors, now under development, may help to meet the demand for energy power of both developed and developing countries as well as for district heating, desalination and for process heat.The paper gives historical examples that illustrate the steps pursued from first research nuclear reactors to present advanced power reactors. Emphasis was laid upon the fact that the progress is due to the great discoveries of the nuclear scientists using the technological transfer.

  10. Proposed power upgrade of the Hot Fuel Examination Facility's neutron radiography reactor. [NRAD reactor

    SciTech Connect

    Pruett, D.P.; Richards, W.J.; Heidel, C.C.

    1984-01-01

    The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both non-destructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the non-destructive examination techniques utilized at HFEF is neutron radiography. Neutron radiography is provided by the NRAD reactor facility, which is located beneath the HFEF hot cell. The NRAD reactor is a TRIGA reactor and is operated at a steady state power level of 250 kW solely for neutron radiography and the development of radiography techniques. When the NRAD facility was designed and constructed, an operating power level of 250 kW was considered to be adequate for obtaining radiographs of the type of specimens envisaged at that time. A typical radiograph required approximately a twenty-minute exposure time. Specimens were typically single fuel rods placed in an aluminum tray. Since that time, however, several things have occurred that have tended to increase radiography exposure times to as much as 90 minutes each. In order to decrease exposure times, the reactor power level is to be increased from 250 kw to 1 MW. This increase in power will necessitate several engineering and design changes. These changes are described.

  11. Power flow control using distributed saturable reactors

    DOEpatents

    Dimitrovski, Aleksandar D.

    2016-02-13

    A magnetic amplifier includes a saturable core having a plurality of legs. Control windings wound around separate legs are spaced apart from each other and connected in series in an anti-symmetric relation. The control windings are configured in such a way that a biasing magnetic flux arising from a control current flowing through one of the plurality of control windings is substantially equal to the biasing magnetic flux flowing into a second of the plurality of control windings. The flow of the control current through each of the plurality of control windings changes the reactance of the saturable core reactor by driving those portions of the saturable core that convey the biasing magnetic flux in the saturable core into saturation. The phasing of the control winding limits a voltage induced in the plurality of control windings caused by a magnetic flux passing around a portion of the saturable core.

  12. Background radiation measurements at high power research reactors

    NASA Astrophysics Data System (ADS)

    Ashenfelter, J.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yeh, M.; Yen, Y.-R.; Zhang, C.; Zhang, X.

    2016-01-01

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  13. Background radiation measurements at high power research reactors

    DOE PAGESBeta

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; et al

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the backgroundmore » fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.« less

  14. Background radiation measurements at high power research reactors

    SciTech Connect

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yen, Y. -R.; Zhang, C.; Zhang, X.

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  15. The neutronics studies of fusion fission hybrid power reactor

    SciTech Connect

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

    2012-06-19

    In this paper, a series of neutronics analysis of hybrid power reactor is proposed. The ideas of loading different fuels in a modular-type fission blanket is analyzed, fitting different level of fusion developments, i.e., the current experimental power output, the level can be obtained in the coming future and the high-power fusion reactor like ITER. The energy multiplication of fission blankets and tritium breeding ratio are evaluated as the criterion of design. The analysis is implemented based on the D-type simplified model, aiming to find a feasible 1000MWe hybrid power reactor for 5 years' lifetime. Three patterns are analyzed: 1) for the low fusion power, the reprocessed fuel is chosen. The fuel with high plutonium content is loaded to achieve large energy multiplication. 2) For the middle fusion power, the spent fuel from PWRs can be used to realize about 30 times energy multiplication. 3) For the high fusion power, the natural uranium can be directly used and about 10 times energy multiplication can be achieved.

  16. Identifying and bounding uncertainties in nuclear reactor thermal power calculations

    SciTech Connect

    Phillips, J.; Hauser, E.; Estrada, H.

    2012-07-01

    Determination of the thermal power generated in the reactor core of a nuclear power plant is a critical element in the safe and economic operation of the plant. Direct measurement of the reactor core thermal power is made using neutron flux instrumentation; however, this instrumentation requires frequent calibration due to changes in the measured flux caused by fuel burn-up, flux pattern changes, and instrumentation drift. To calibrate the nuclear instruments, steam plant calorimetry, a process of performing a heat balance around the nuclear steam supply system, is used. There are four basic elements involved in the calculation of thermal power based on steam plant calorimetry: The mass flow of the feedwater from the power conversion system, the specific enthalpy of that feedwater, the specific enthalpy of the steam delivered to the power conversion system, and other cycle gains and losses. Of these elements, the accuracy of the feedwater mass flow and the feedwater enthalpy, as determined from its temperature and pressure, are typically the largest contributors to the calorimetric calculation uncertainty. Historically, plants have been required to include a margin of 2% in the calculation of the reactor thermal power for the licensed maximum plant output to account for instrumentation uncertainty. The margin is intended to ensure a cushion between operating power and the power for which safety analyses are performed. Use of approved chordal ultrasonic transit-time technology to make the feedwater flow and temperature measurements (in place of traditional differential-pressure- based instruments and resistance temperature detectors [RTDs]) allows for nuclear plant thermal power calculations accurate to 0.3%-0.4% of plant rated power. This improvement in measurement accuracy has allowed many plant operators in the U.S. and around the world to increase plant power output through Measurement Uncertainty Recapture (MUR) up-rates of up to 1.7% of rated power, while also

  17. Safety status of space radioisotope and reactor power sources

    NASA Technical Reports Server (NTRS)

    Bennett, Gary L.

    1990-01-01

    The current overall safety criterion for both radioisotope and reactor power sources is containment or immobilization in the case of a reentry accident. In addition, reactors are designed to remain subcritical under conditions of land impact or water immersion. A very extensive safety test and analysis program was completed on the radioisotope thermoelectric generators (RTGs) in use on the Galileo spacecraft and planned for use on the Ulysses spacecraft. The results of this work show that the RTGs will pose little or no risk for any credible accident. The SP-100 space nuclear reactor program has begun addressing its safety criteria, and the design is planned to be such as to ensure meeting the various safety criteria. Preliminary mission risk analyses on SP-100 show the expected value population dose from postulated accidents on the reference mission to be very small. It is concluded that the current US nuclear power sources are the safest flown.

  18. Design considerations for an inertial confinement fusion reactor power plant

    SciTech Connect

    Massey, J.V.; Simpson, J.E.

    1981-08-10

    To further define the engineering and economic concerns for inertial confinement fusion reactors (ICR's), a conceptual design study was performed by Bechtel Group Incorporated under the direction of Lawrence Livermore National Laboratory (LLNL). The study examined alternatives to the LLNL HYLIFE concept and expanded the previous balance of plant design to incorporate information from recent liquid metal cooled fast breeder reactor (LMFBR) power plant studies. The majority of the effort was to incorporate present laser and target physics models into a reactor design with a low coolant flowrate and a high driver repetition rate. An example of such a design is the LLNL JADE concept. In addition to producing a power plant design for LLNL using the JADE example, Bechtel has also examined the applicability of the EAGLE (Energy Absorbing Gas Lithium Ejector) concept.

  19. Space nuclear reactor power-perception, requirements, and mission integration

    NASA Astrophysics Data System (ADS)

    Isenberg, Lon; Martin, C. R.

    The authors describe the opportunities for the military and the civil sector given a successful development of a space reactor power system. The centerpiece is the SP-100 program. Numerous studies have identified serious design limitations for military missions. These same studies suggest that thermionic technology has great promise for military missions. By way of highlighting some of the concerns over the current SP-100 program, the authors focus on some of the technology issues. Recently, as a result of two military studies, the military and the Department of Energy (DOE) have proposed a memorandum of understanding to cosponsor a second reactor development program based on thermionic technology. The authors believe the DOE should strongly endorse joint development of thermionics with the military. Additionally, the reactor power source should be closely integrated with a particular mission.

  20. Estimates of power requirements for a Manned Mars Rover powered by a nuclear reactor

    NASA Technical Reports Server (NTRS)

    Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

    1991-01-01

    This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are meet using an SP-100 type reactor. The primary electric power needs, which include 30-kW(e) net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle using He/Xe as the working fluid. The specific mass of the nuclear reactor power system, including a man-rated radiation shield, ranged from 150-kg/kW(e) to 190-kg/KW(e) and the total mass of the Rover vehicle varied depend upon the cruising speed.

  1. Estimates of power requirements for a Manned Mars Rover powered by a nuclear reactor

    NASA Astrophysics Data System (ADS)

    Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

    This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are meet using an SP-100 type reactor. The primary electric power needs, which include 30-kW(e) net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle using He/Xe as the working fluid. The specific mass of the nuclear reactor power system, including a man-rated radiation shield, ranged from 150-kg/kW(e) to 190-kg/KW(e) and the total mass of the Rover vehicle varied depend upon the cruising speed.

  2. Estimates of power requirements for a manned Mars rover powered by a nuclear reactor

    NASA Astrophysics Data System (ADS)

    Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

    1991-01-01

    This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are met using an SP-100 type reactor. The primary electric power needs, which include 30-kWe net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine (FPSE) yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle (CBC) using He/Xe as the working fluid. The specific mass of the nuclear reactor power systrem, including a man-rated radiation shield, ranged from 150-kg/kWe to 190-kg/kWe and the total mass of the Rover vehicle varied depend upon the cruising speed.

  3. Systems aspects of a space nuclear reactor power system

    NASA Technical Reports Server (NTRS)

    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.

  4. Fuel element concept for long life high power nuclear reactors

    NASA Technical Reports Server (NTRS)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  5. 78 FR 64028 - Decommissioning of Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... Register on February 14, 2012 (77 FR 8902), for a 60-day public comment period. The public comment period... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission....

  6. High density operation for reactor-relevant power exhaust

    NASA Astrophysics Data System (ADS)

    Wischmeier, M.

    2015-08-01

    With increasing size of a tokamak device and associated fusion power gain an increasing power flux density towards the divertor needs to be handled. A solution for handling this power flux is crucial for a safe and economic operation. Using purely geometric arguments in an ITER-like divertor this power flux can be reduced by approximately a factor 100. Based on a conservative extrapolation of current technology for an integrated engineering approach to remove power deposited on plasma facing components a further reduction of the power flux density via volumetric processes in the plasma by up to a factor of 50 is required. Our current ability to interpret existing power exhaust scenarios using numerical transport codes is analyzed and an operational scenario as a potential solution for ITER like divertors under high density and highly radiating reactor-relevant conditions is presented. Alternative concepts for risk mitigation as well as strategies for moving forward are outlined.

  7. A gas-cooled reactor surface power system

    NASA Astrophysics Data System (ADS)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  8. A Gas-Cooled Reactor Surface Power System

    SciTech Connect

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  9. A gas-cooled reactor surface power system

    SciTech Connect

    Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1{percent}Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. {copyright} {ital 1999 American Institute of Physics.}

  10. A gas-cooled reactor surface power system

    SciTech Connect

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-22

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  11. Operating margin of Soviet RBMK-1000 nuclear power reactors

    SciTech Connect

    Adams, J.M.; Robinson, G.E. . Dept. of Nuclear Engineering); Hochreiter, L.E. )

    1991-12-01

    This paper reports on a coupled thermal- hydraulic analysis that is performed for the Soviet-designed RBMK-1000 nuclear power reactor to assess the operating margin to critical heat flux (CHF); the Chernobyl-4 reactor serves as the principal model for this study. Calculations are performed using a simplified subchannel analysis. The overall analysis involves an iterative search to determine the individual subchannel flow rates, and a boiling transition analysis is performed to obtain a measure of the core operating margin. The operating margin is determined via two distinct methods. The first involves a calculation of the core critical power ratio (CPR) using an empirically derived correlation that the Soviets developed expressly for the RBMK-1000. Additionally, various subchannel CHF correlations typical of those used in the design of nuclear-powered reactors in the United States are also employed. When the Soviet critical power correlation is used, the calculations carried out for both normal operating and reference overpower conditions result in CPRs of 1.115 and 1.019, respectively. In most cases, the subchannel CHF correlations indicate that additional operating margin over that calculated by the Soviet critical power correlation exists for this design.

  12. Gravity Scaling of a Power Reactor Water Shield

    SciTech Connect

    Reid, Robert S.; Pearson, J. Boise

    2008-01-21

    Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa{sup n}. These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

  13. Closed Brayton cycle power conversion systems for nuclear reactors :

    SciTech Connect

    Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.; Sanchez, Travis

    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

  14. Prompt gamma radiation as a new tool to measure reactor power

    NASA Astrophysics Data System (ADS)

    jalali, Majid; Abdi, Mohammad Reza; davati, Mojtaba Mostajabod

    2013-10-01

    A new method, based on reactor prompt gamma radiation detection, for reactor power measurement is introduced and validated. To verify, the ex-core gamma radiation spectrum from the Iranian Heavy Water Zero Power Reactor (HWZPR) were measured by HPGe and NaI detectors each suitably positioned. The collective prompt gamma count rates for all or for a portion of each of 2″×2″ NaI detector spectra were obtained for seven power level readings from calibrated reactor power monitors. A good linear behavior was found between gamma count rate and reactor power. The method of calibrated prompt gamma reactor power determination is a stable and reliable tool, on-line, sensitive to sudden variation of power, working in pulse mode, increasing redundancy and diversity and so improving the reactor safety. The prompt gamma counting system can be adopted and installed in other nuclear reactors to measure power.

  15. Technological implications of SNAP reactor power system development on future space nuclear power systems

    SciTech Connect

    Anderson, R.V.

    1982-11-16

    Nuclear reactor systems are one method of satisfying space mission power needs. The development of such systems must proceed on a path consistent with mission needs and schedules. This path, or technology roadmap, starts from the power system technology data base available today. Much of this data base was established during the 1960s and early 1970s, when government and industry developed space nuclear reactor systems for steady-state power and propulsion. One of the largest development programs was the Systems for Nuclear Auxiliary Power (SNAP) Program. By the early 1970s, a technology base had evolved from this program at the system, subsystem, and component levels. There are many implications of this technology base on future reactor power systems. A review of this base highlights the need for performing a power system technology and mission overview study. Such a study is currently being performed by Rockwell's Energy Systems Group for the Department of Energy and will assess power system capabilities versus mission needs, considering development, schedule, and cost implications. The end product of the study will be a technology roadmap to guide reactor power system development.

  16. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a...

  17. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a...

  18. 77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-15

    ... COMMISSION Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors AGENCY... ``Decommissioning of Nuclear Power Reactors.'' This guide describes a method NRC considers acceptable for use in... Revision 1 of Regulatory Guide 1.184, ``Decommissioning of Nuclear Power Reactors,'' dated July 2000....

  19. Neutron Spectra and Dose Equivalent Inside Nuclear Power Reactor Containment

    SciTech Connect

    Aldrich, J. M.

    1981-08-01

    This study was conducted to determine absorbed dose, dose-equivalent rates, and neutron spectra inside containment at nuclear power plants. We gratefully acknowledge funding support by the Nuclear Regulatory Commission. The purpose of this study is: 1) measure dose-equivalent rates with various commercial types of rem meters, such as the Snoopy and Rascal, and neutron absorbed dose rates with a tissue-equivalent proportional counter 2) determine neutron spectra using the multi sphere or Bonner sphere technique and a helium-3 spectrometer 3) compare several types of personnel neutron dosimeter responses such as NTA film, polycarbonates, TLD albedo, and a recently introduced proton recoil track etch dosimeter, and CR-39. These measurements were made inside containments of pressurized water reactors (PWRs) and outside containment penetrations of boiling water reactors (BWRs) operating at full power. The neutron spectral information, absorbed dose. and dose-equivalent measurements are needed for proper interpretation of instrument and personnel dosimeter responses.

  20. Power Conversion Study for High Temperature Gas-Cooled Reactors

    SciTech Connect

    Chang Oh; Richard Moore; Robert Barner

    2005-05-01

    The Idaho National Laboratory (INL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. There are some technical issues to be resolved before the selection of the final design of the high temperature gascooled reactor, called as a Next Generation Nuclear Plant (NGNP), which is supposed to be built at the INEEL by year 2017. The technical issues are the selection of the working fluid, direct vs. indirect cycle, power cycle type, the optimized design in terms of a number of intercoolers, and others. In this paper, we investigated a number of working fluids for the power conversion loop, direct versus indirect cycle, the effect of intercoolers, and other thermal hydraulics issues. However, in this paper, we present part of the results we have obtained. HYSYS computer code was used along with a computer model developed using Visual Basic computer language.

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

    SciTech Connect

    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.

  2. Gravity Scaling of a Power Reactor Water Shield

    NASA Technical Reports Server (NTRS)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for use on initial lunar surface power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxiliary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2007). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n). These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

  3. Autonomous Control Capabilities for Space Reactor Power Systems

    SciTech Connect

    Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

    2004-02-04

    The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

  4. Autonomous Control Capabilities for Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

    2004-02-01

    The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

  5. Nuclear safety as applied to space power reactor systems

    SciTech Connect

    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.

  6. Test plan for the Parallex CANDU-MOX irradiation

    SciTech Connect

    Copeland, G.L.

    1997-06-01

    One of several options being considered by the United States and the Russian Federation for the disposition of excess plutonium from dismantled weapons is to convert it to mixed-oxide (MOX) fuel for use in Canadian uranium-deuterium (CANDU) reactors. This report describes an irradiation test demonstrating the feasibility of this concept with laboratory quantities of MOX fuel placed in the pressurized loops of the National Research Universal test reactor at the Atomic Energy of Canada, Ltd., Chalk River Laboratories. The objective of the Parallex (for parallel experiment) test is to simultaneously test laboratory-produced quantities of US and R.F. MOX fuel in a test reactor under heat generation rates representing those expected in the CANDU reactors. The MOX fuel will be produced with plutonium from disassembled weapons at the Los Alamos National Laboratory in the United States and at the Bochvar Institute in the Russian Federation. Thus, the test will serve to demonstrate the accomplishment of many parts of the disposition mission: disassembly of weapons, conversion of the plutonium to oxide, fabrication of MOX fuel, assembly of fuel elements and bundles, shipment to a reactor, irradiation, and finally, storage of the spent fuel elements awaiting eventual disposition in a geologic repository in Canada.

  7. Design Concept for a Nuclear Reactor-Powered Mars Rover

    NASA Astrophysics Data System (ADS)

    Elliott, John O.; Lipinski, Ronald J.; Poston, David I.

    2003-01-01

    A study was recently carried out by a team from JPL and the DOE to investigate the utility of a DOE-developed 3 kWe surface fission power system for Mars missions. The team was originally tasked to perform a study to evaluate the usefulness and feasibility of incorporation of such a power system into a landed mission. In the course of the study it became clear that the application of such a power system was enabling to a wide variety of potential missions. Of these, two missions were developed, one for a stationary lander and one for a reactor-powered rover. This paper discusses the design of the rover mission, which was developed around the concept of incorporating the fission power system directly into a large rover chassis to provide high power, long range traverse capability. The rover design is based on a minimum extrapolation of technology, and adapts existing concepts developed at JPL for the 2009 Mars Science Laboratory (MSL) rover, lander and EDL systems. The small size of the reactor allowed its incorporation directly into an existing large MSL rover chassis design, allowing direct use of MSL aeroshell and pallet lander elements, beefed up to support the significantly greater mass involved in the nuclear power system and its associated shielding. This paper describes the unique design challenges encountered in the development of this mission architecture and incorporation of the fission power system in the rover, and presents a detailed description of the final design of this innovative concept for providing long range, long duration mobility on Mars.

  8. Supercritical Water Reactor Cycle for Medium Power Applications

    SciTech Connect

    BD Middleton; J Buongiorno

    2007-04-25

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency {ge}20%; Steam turbine outlet quality {ge}90%; and Pumping power {le}2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

  9. Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

    SciTech Connect

    M. L. Grossbeck J-P.A. Renier Tim Bigelow

    2003-09-30

    Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility and cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing fuel life in commercial reactors, in particular Generation-IV reactors. The project also investigated various burnable poison configurations, and studied incorporation of metallic burnable poisons into fuel cladding.

  10. Fast reactor power plant design having heat pipe heat exchanger

    DOEpatents

    Huebotter, P.R.; McLennan, G.A.

    1984-08-30

    The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

  11. Fast reactor power plant design having heat pipe heat exchanger

    DOEpatents

    Huebotter, Paul R.; McLennan, George A.

    1985-01-01

    The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

  12. Issues in the flight qualification of a space power reactor

    SciTech Connect

    Polansky, G.F.; Schmidt, G.L.; Voss, S.S.; Reynolds, E.L.

    1994-10-01

    This paper presents an overview of the Nuclear Electric Propulsion Space Test Program (NEPSTP). The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between US and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year.

  13. Gas Core Reactor with Magnetohydrodynamic Power System and Cascading Power Cycle

    SciTech Connect

    Smith, Blair M.; Anghaie, Samim

    2004-03-15

    The U.S. Department of Energy initiative Generation IV aim is to produce an entire nuclear energy production system with next-generation features for certification before 2030. A Generation IV-capable system must have superior sustainability, safety and reliability, and economic cost advantages in comparison with third generation light water reactors (LWRs). A gas core reactor (GCR) with magnetohydrodynamic (MHD) power converter and cascading power cycle forms the basis for a Generation IV concept that is expected to set the upper performance limits in sustainability and power conversion efficiency among all existing and proposed fission powered systems. A gaseous core reactor delivering thousands of megawatt fission power acts as the heat source for a high-temperature MHD power converter. A uranium tetrafluoride fuel mix, with {approx}95% mol fraction helium gas, provides a stable working fluid for the primary MHD Brayton cycle. The hot working fluid exiting a topping cycle MHD generator has sufficient heat to drive a conventional helium Brayton cycle with 35% thermal efficiency as well as a superheated steam Rankine cycle, with up to 40% efficiency, which recovers the waste heat from the intermediate Brayton cycle. A combined cycle efficiency of close to 70% can be achieved with only a modest MHD topping cycle efficiency. The high-temperature direct-energy conversion capability of an MHD dynamo combined with an already sophisticated steam-powered turbine industry knowledge base allows the cascading cycle design to achieve breakthrough first-law energy efficiencies previously unheard of in the nuclear power industry. Although simple in concept, the gas core reactor design has not achieved the state of technological maturity that established high-temperature gas-cooled reactors and high-temperature molten salt core reactors have pioneered. However, the GCR-MHD concept has considerable promise; for example, like molten salt reactors the fuel is continuously cycled

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

    SciTech Connect

    Cooke, Conrad; Spann, Holger

    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 Plates to

  15. PR-EDB: Power Reactor Embrittlement Database - Version 3

    SciTech Connect

    Wang, Jy-An John; Subramani, Ranjit

    2008-03-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  16. Analyse du transfert de chaleur et de la perte de pression pour des ecoulements supercritiques dans le reacteur CANDU-SCWR

    NASA Astrophysics Data System (ADS)

    Zoghlami, Sarra

    channel, some geometrical simplifications are made. In fact, we assumed that the coolant flow in the fuel channel is represented by the flow around the fuel rod, bounded by the first crown. This simplified model was used for neutronic/thermal-hydraulic coupled calculation performed with neutronic codes DRAGON/DONJON (Varin et al., 2005), to analyse the thermal-hydraulic behavior of the fuel channel in CANDU-SCWR. We observed that the coolant density at the fuel rod external surface, at the sixth fuel bundle is 3.5 times lower than the average coolant density in the fuel channel. This puts into question the assumption of considering the supercritical water flow as an homogeneous flow and the ability to build a supercritical water nuclear reactor in CANDU type, i.e., with horizontal fuel channels. In order to validate ARTHUR &barbelow;SCWR code, we compared our results to SUBCHAN code (Jiang et al., 2009), which is a thermal-hydraulic code developed to analyze CANDU-SCWR. Both codes give the same shapes and orders of magnitude for the coolant average temperature and the cladding-surface temperature. The axial distribution of the centerline temperature in the fuel rod is different. This is due to the fact that the calculations performed by the SUBCHAN code are not coupled to a neutronic code. For this reason, the thermal power distributions differs in the two codes. The variation of the mass flow influences the forced convective heat transfer, so, the distribution of thermal-physical properties in the channel. In fact, if the mass flow is reduced by 50% compared to the nominal mass flow rate, following a pump failure, the external fuel rod surface temperature exceeds the melting point, which is between 1400°C and 1455°C. This phenomenon may results in the radioactive contamination of the environment.

  17. Small space reactor power systems for unmanned solar system exploration missions

    SciTech Connect

    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.

  18. Small space reactor power systems for unmanned solar system exploration missions

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

    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.

  19. Gas Core Reactor-MHD Power System with Cascading Power Cycle

    SciTech Connect

    Smith, Blair M.; Anghaie, Samim; Knight, Travis W.

    2002-07-01

    The US Department of Energy initiative Gen-IV aim is to produce an entire nuclear energy production system with next generation features for certification before 2030. A Generation 4 capable system must have superior sustainability, safety and reliability, and economic cost advantages in comparison with third generation light water reactors. A gas core reactor (GCR) with magnetohydrodynamic (MHD) power converter and cascading power cycle forms the basis for a Generation IV concept that is expected to set the upper performance limits in sustainability and power conversion efficiency among all existing and proposed fission powered systems. A gaseous core reactor delivering 1000's MW fission power acts as the heat source for a high temperature magnetohydrodynamic power converter. A uranium tetrafluoride fuel mix, with {approx}95% mole fraction helium gas, provides a stable working fluid for the primary MHD-Brayton cycle. A helium Brayton cycle extracts waste heat from the MHD generator with about 20% energy efficiency, but the low temperature side is still hot enough ({approx}1600 K) to drive a second conventional helium Brayton cycle with about 35% efficiency. There is enough heat at the low temperature side of the He-Brayton cycle to generate steam, and so another heat recovery cycle can be added, this time a Rankine steam cycle with up to 40% efficiency. The proof of concept does not require a tremendously efficient (first law) MHD cycle, the high temperature direct energy conversion capability of an MHD dynamo, combined with already sophisticated steam powered turbine industry knowledge base allows the cascading cycle design to achieve break-through first law energy efficiencies previously unheard of in the nuclear power industry. Although simple in concept, the gas core reactor design has not achieved the state of technological maturity that, say, molten salt or high-temperature gas-cooled reactors have pioneered. However, even on paper the GCR-MHD concept holds

  20. Computer study of emergency shutdowns of a 60-kilowatt reactor Brayton space power system

    NASA Technical Reports Server (NTRS)

    Tew, R. C.; Jefferies, K. S.

    1974-01-01

    A digital computer study of emergency shutdowns of a 60-kWe reactor Brayton power system was conducted. Malfunctions considered were (1) loss of reactor coolant flow, (2) loss of Brayton system gas flow, (3)turbine overspeed, and (4) a reactivity insertion error. Loss of reactor coolant flow was the most serious malfunction for the reactor. Methods for moderating the reactor transients due to this malfunction are considered.

  1. System aspects of a Space Nuclear Reactor Power System

    SciTech Connect

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

    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, altitude 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. The power system design evolved during the study and has continued to evolve; the current design differs somewhat from that examined in this paper.

  2. Medium Power Lead Alloy Fast Reactor Balance of Plant Options

    SciTech Connect

    Vaclav Dosta; Pavel Hejzlar; Neil E. Todreas; Jacopo Buongiorno

    2004-09-01

    Proper selection of the power conversion cycle is a very important step in the design of a nuclear reactor. Due to the higher core outlet temperature (~550°C) compared to that of light water reactors (~300°C), a wide portfolio of power cycles is available for the lead alloy fast reactor (LFR). Comparison of the following cycles for the LFR was performed: superheated steam (direct and indirect), supercritical steam, helium Brayton, and supercritical CO2 (S-CO2) recompression. Heat transfer from primary to secondary coolant was first analyzed and then the steam generators or heat exchangers were designed. The direct generation of steam in the lead alloy coolant was also evaluated. The resulting temperatures of the secondary fluids are in the range of 530-545°C, dictated by the fixed space available for the heat exchangers in the reactor vessel. For the direct steam generation situation, the temperature is 312°C. Optimization of each power cycle was carried out, yielding net plant efficiency of around 40% for the superheated steam cycle while the supercritical steam and S-CO2 cycles achieved net plant efficiency of 41%. The cycles were then compared based on their net plant efficiency and potential for low capital cost. The superheated steam cycle is a very good candidate cycle given its reasonably high net plant efficiency and ease of implementation based on the extensive knowledge and operating experience with this cycle. Although the supercritical steam cycle net plant efficiency is slightly better than that of the superheated steam cycle, its high complexity and high pressure result in higher capital cost, negatively affecting plant economics. The helium Brayton cycle achieves low net plant efficiency due to the low lead alloy core outlet temperature, and therefore, even though it is a simpler cycle than the steam cycles, its performance is mediocre in this application. The prime candidate, however, appears to be the S-CO2 recompression cycle, because it

  3. Calculation of the Local Neutronic Parameters for CANDU Fuel Bundles Using Transport Methods

    SciTech Connect

    Balaceanu, Victoria; Rizoiu, Andrei; Hristea, Viorel

    2006-07-01

    For a realistic neutronic evaluation of the CANDU reactor core it is important to accurately perform the local neutronic parameters (i.e. multigroup macroscopic cross sections for the core materials) calculation. This means using codes that allow a good geometric representation of the CANDU fuel bundle and then solving the transport equation. The paper reported here intends to study in detail the local behavior for two types of CANDU fuel, NU{sub 3}7 (Natural Uranium, 37 elements) and SEU{sub 4}3 (Slightly Enriched Uranium, 43 elements, with 1.1 wt% enrichment). The considered fuel types represent fresh and used bundles. The two types of CANDU super-cells are reference NU{sub 3}7, perturbed NU{sub 3}7, reference SEU{sub 4}3 and perturbed SEU{sub 4}3. The perturbed super-cells contain a Mechanical Control Absorber (a very strong reactivity device). For reaching the proposed objective a methodology is used based on WIMS and PIJXYZ codes. WIMS is a standard lattice-cell code, based on transport theory and it is used for producing fuel cell multigroup macroscopic cross sections. For obtaining the fine local neutronic parameters in the CANDU super-cells (k-eff values, local MCA reactivity worth, flux distributions and reaction rates), the PIJXYZ code is used. PIJXYZ is a 3D integral transport code using the first collision probability method and it has been developed for CANDU cell geometry. It is consistent with WIMS lattice-cell calculations and allows a good geometrical representation of the CANDU bundle in three dimensions. The analysis of the neutronic parameters consists of comparing the obtained results with the similar results calculated with the DRAGON code. This comparison shows a good agreement between these results. (authors)

  4. Validation of WIMS-IST for CANDU R-type lattices

    SciTech Connect

    Bromley, B. P.; Davis, R.

    2006-07-01

    Prior validation studies of 28-element natural uranium (28-NU) CANDU R-type fuel bundles using the WIMS-IST lattice physics code had demonstrated a bias in the calculation of the coolant void reactivity (CVR) of approximately +0.5 to +0.6 mk (1 mk =100 pcm = 0.001 {Delta}k/k). However, these validation studies were performed using experimental data for 28-element bundles with pressure tubes that were smaller than standard CANDU-type pressure tubes, giving a smaller coolant volume, and a modified neutron energy spectrum. Validation studies performed with 37-element and 43-element fuel bundles with a CANDU-type lattice pitch and pressure tube had shown a CVR bias of {approx} 1.7 to 1.9 mk. It was believed that the CVR bias for a 28-element bundle would be closer to this range of values if a standard CANDU pressure tube diameter were used The objective of this study was to confirm this hypothesis, that using a larger CANDU-standard pressure tube would give a larger CVR bias for a 28-NU fuel bundle, as computed by WIMS-IST in comparison to experimental measurements of critical buckling. Thus, new critical-height and flux-map measurements were performed in substitution experiments in the ZED-2 research reactor to determine the pure critical lattice buckling for 28-element fuel with standard-size CANDU pressure tubes. The derived buckling from these experiments were used in WIMS-IST transport calculations to determine the effective multiplication factors for cooled and voided lattices and hence the bias in the CVR. Calculation results demonstrated that the CVR bias for the 28-NU was {approx} 1.7 mk {+-} 0.42 mk, which is consistent with the results for 37-element and 43-element CANDU-type lattices. (authors)

  5. A novel technique to remove deuterium from CANDU pressure tubes

    NASA Astrophysics Data System (ADS)

    Qin, Z.; Zhang, C.-S.; Griffiths, K.; Norton, P. R.

    2001-10-01

    Deuterium ingress into the pressure tubes of a Canada deuterium uranium (CANDU) nuclear reactor can cause the pressure tubes to crack prematurely. A novel technique, based on the rapid diffusion of deuterium in zirconium alloys, and subsequent preferential segregation of deuterium at the surface, has been developed to remove dissolved deuterium. This technique involves a simple continuous plasma treatment of the surface of a pressure tube, and can remove as much as 70% of the dissolved deuterium from the entire wall thickness of a pressure tube in realistic time scales. The proposed technique has considerable economic incentive: it may extend the life of pressure tubes without channel replacement.

  6. A preliminary investigation of the Topaz II reactor as a lunar surface power supply

    SciTech Connect

    Polansky, G.F.; Houts, M.G.

    1995-12-31

    Reactor power supplies offer many attractive characteristics for lunar surface applications. The Topaz II reactor resulted from an extensive development program in the former Soviet Union. Flight quality reactor units remain from this program and are currently under evaluation in the United States. This paper examines the potential for applying the Topaz II, originally developed to provide spacecraft power, as a lunar surface power supply.

  7. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  8. Reactor/Brayton power systems for nuclear electric spacecraft

    NASA Technical Reports Server (NTRS)

    Layton, J. P.

    1980-01-01

    Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. This vehicle would be capable of performing detailed exploration of the outer planets of the solar system during the remainder of this century. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system. The results have shown that the performance is very competitive and that a 400 kWe space power system is dimensionally compatible with a single Space Shuttle launch. Performance parameters of system mass and radiator area were determined for systems from 100 to 1000 kWe. A 400 kWe reference system received primary attention. The components of this system were defined and a conceptual layout was developed with encouraging results. The preliminary mass determination for the complete power system was very close to the desired goal of 20 kg/kWe. Use of more advanced technology (higher turbine inlet temperature) will substantially improve system performance characteristics.

  9. ENHANCING ADVANCED CANDU PROLIFERATION RESISTANCE FUEL WITH MINOR ACTINIDES

    SciTech Connect

    Gray S. Chang

    2010-05-01

    The advanced nuclear system will significantly advance the science and technology of nuclear energy systems and to enhance the spent fuel proliferation resistance. Minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs can play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In this work, an Advanced CANDU Reactor (ACR) fuel unit lattice cell model with 43 UO2 fuel rods will be used to investigate the effectiveness of a Minor Actinide Reduction Approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. The main MARA objective is to increase the 238Pu / Pu isotope ratio by using the transuranic nuclides (237Np and 241Am) in the high burnup fuel and thereby increase the proliferation resistance even for a very low fuel burnup. As a result, MARA is a very effective approach to enhance the proliferation resistance for the on power refueling ACR system nuclear fuel. The MA transmutation characteristics at different MA loadings were compared and their impact on neutronics criticality assessed. The concept of MARA, significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance.

  10. Power control of SAFE reactor using fuzzy logic

    NASA Astrophysics Data System (ADS)

    Irvine, Claude

    2002-01-01

    Controlling the 100 kW SAFE (Safe Affordable Fission Engine) reactor consists of design and implementation of a fuzzy logic process control system to regulate dynamic variables related to nuclear system power. The first phase of development concentrates primarily on system power startup and regulation, maintaining core temperature equilibrium, and power profile matching. This paper discusses the experimental work performed in those areas. Nuclear core power from the fuel elements is simulated using resistive heating elements while heat rejection is processed by a series of heat pipes. Both axial and radial nuclear power distributions are determined from neuronic modeling codes. The axial temperature profile of the simulated core is matched to the nuclear power profile by varying the resistance of the heating elements. The SAFE model establishes radial temperature profile equivalence by establishing 32 control zones as the nodal coordinates. Control features also allow for slow warm up, since complete shutoff can occur in the heat pipes if heat-source temperatures drop/rise below a certain minimum value, depending on the specific fluid and gas combination in the heat pipe. The entire system is expected to be self-adaptive, i.e., capable of responding to long-range changes in the space environment. Particular attention in the development of the fuzzy logic algorithm shall ensure that the system process remains at set point, virtually eliminating overshoot on start-up and during in-process disturbances. The controller design will withstand harsh environments and applications where it might come in contact with water, corrosive chemicals, radiation fields, etc. .

  11. Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

    SciTech Connect

    Jung, Y. S.; Joo, H. G.; Yoon, J. I.

    2013-07-01

    The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

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

    NASA Technical Reports Server (NTRS)

    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.

  13. System simulation of a multicell thermionic space power reactor

    NASA Astrophysics Data System (ADS)

    von Arx, Alan Vincent

    For many years, thermionic power has been considered for space application. The prominent feature of the power conversion system is that there are no moving parts. Although designs have been developed by various organizations, no comprehensive system models are known to exist which can simulate transient behavior of a multicell design nor is there a method to directly couple these models to other codes that can calculate variations in reactivity. Thus, a procedure has been developed to couple the performance calculations of a space nuclear reactor thermal/hydraulics code with a neutron diffusion code to analyze temperature feedback. Thermionic power is based on the thermionic emissions principle where free electrons in a conductor have sufficient energy to escape the surface. Kinetic energy is given to the electrons by heating the conductor. Specifically, a 48 kWe thermionic power converter system model has been developed and used to model startup and other transients. Less than 10% of the fuel heat is converted to electricity, and the rest is rejected to space via a heat pipe radiator. An electromagnetic pump circulates the liquid metal coolant. First, a startup transient model was developed which showed stable operation through ignition of the Thermionic Fuel Elements (TFEs) and thawing of the radiator heat pipes. Also, the model's capability was expanded to include two-phase heat transfer to model boiling using coupled mass and thermal energy conservation equations. The next step incorporated effects of reactivity feedback---showing that various mechanisms will prevent power and temperature run-up for a flow reduction scenario where the reactor control systems fail to respond. In particular, the Doppler effect was shown to counter a positive worth due to partial core voiding although steps must be taken to preclude film boiling in that high superheats will result in TFE failures. Finally, analysis of the core grid spacer location suggests it should be located at

  14. Current Trends of Blanket Research and Deveopment in Japan 3.Blanket Designs in Fusion Power Reactors

    NASA Astrophysics Data System (ADS)

    Sagara, Akio; Enoeda, Mikio; Nishio, Satoshi; Kozaki, Yasuji

    The main functions of the blanket in fusion power reactors are basically independent of the type of magnetic fusion reactor (tokamak, helical, etc.) and inertia fusion. However, from technical point of view, many candidate designs of blanket have been proposed depending on the particular reactor concepts. Their main features are characterized for the recent typical designs, and key issues are defined.

  15. Thermoelectric converter for SP-100 space reactor power system

    NASA Technical Reports Server (NTRS)

    Terrill, W. R.; Haley, V. F.

    1986-01-01

    Conductively coupling the thermoelectric converter to the heat source and the radiator maximizes the utilization of the reactor and radiator temperatures and thereby minimizes the power system weight. This paper presents the design for the converter and the individual thermoelectric cells that are the building block modules for the converter. It also summarizes progress on the fabrication of initial cells and the results obtained from the preparation of a manufacturing plan. The design developed for the SP-100 system utilizes thermally conductive compliant pads that can absorb the displacement and distortion caused by the combinations of temperatures and thermal expansion coefficients. The converter and cell designs provided a 100 kWe system which met the system requirements. Initial cells were fabricated and tested.

  16. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    SciTech Connect

    1995-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience.

  17. Method of locating a leaking fuel element in a fast breeder power reactor

    DOEpatents

    Honekamp, John R.; Fryer, Richard M.

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  18. High-intensity power-resolved radiation imaging of an operational nuclear reactor.

    PubMed

    Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  19. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    NASA Astrophysics Data System (ADS)

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-10-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

  20. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    PubMed Central

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  1. Square lattice honeycomb reactor for space power and propulsion

    NASA Astrophysics Data System (ADS)

    Gouw, Reza; Anghaie, Samim

    2000-01-01

    The most recent nuclear design study at the Innovative Nuclear Space Power and Propulsion Institute (INSPI) is the Moderated Square-Lattice Honeycomb (M-SLHC) reactor design utilizing the solid solution of ternary carbide fuels. The reactor is fueled with solid solution of 93% enriched (U,Zr,Nb)C. The square-lattice honeycomb design provides high strength and is amenable to the processing complexities of these ultrahigh temperature fuels. The optimum core configuration requires a balance between high specific impulse and thrust level performance, and maintaining the temperature and strength limits of the fuel. The M-SLHC design is based on a cylindrical core that has critical radius and length of 37 cm and 50 cm, respectively. This design utilized zirconium hydrate to act as moderator. The fuel sub-assemblies are designed as cylindrical tubes with 12 cm in diameter and 10 cm in length. Five fuel subassemblies are stacked up axially to form one complete fuel assembly. These fuel assemblies are then arranged in the circular arrangement to form two fuel regions. The first fuel region consists of six fuel assemblies, and 18 fuel assemblies for the second fuel region. A 10-cm radial beryllium reflector in addition to 10-cm top axial beryllium reflector is used to reduce neutron leakage from the system. To perform nuclear design analysis of the M-SLHC design, a series of neutron transport and diffusion codes are used. To optimize the system design, five axial regions are specified. In each axial region, temperature and fuel density are varied. The axial and radial power distributions for the system are calculated, as well as the axial and radial flux distributions. Temperature coefficients of the system are also calculated. A water submersion accident scenario is also analyzed for these systems. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel, which provides a relatively

  2. Power monitoring in space nuclear reactors using silicon carbide radiation detectors

    NASA Technical Reports Server (NTRS)

    Ruddy, Frank H.; Patel, Jagdish U.; Williams, John G.

    2005-01-01

    Space reactor power monitors based on silicon carbide (SiC) semiconductor neutron detectors are proposed. Detection of fast leakage neutrons using SiC detectors in ex-core locations could be used to determine reactor power: Neutron fluxes, gamma-ray dose rates and ambient temperatures have been calculated as a function of distance from the reactor core, and the feasibility of power monitoring with SiC detectors has been evaluated at several ex-core locations. Arrays of SiC diodes can be configured to provide the required count rates to monitor reactor power from startup to full power Due to their resistance to temperature and the effects of neutron and gamma-ray exposure, SiC detectors can be expected to provide power monitoring information for the fill mission of a space reactor.

  3. Monitoring the thermal power of nuclear reactors with a prototype cubic meter antineutrino detector

    NASA Astrophysics Data System (ADS)

    Bernstein, A.; Bowden, N. S.; Misner, A.; Palmer, T.

    2008-04-01

    In this paper, we estimate how quickly and how precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Our results are obtained from a detector we have deployed and operated at 25m standoff from a reactor core. This prototype can detect a prompt reactor shutdown within 5h and monitor relative thermal power to within 7days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's reactor safeguards regime or other cooperative monitoring regimes.

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

    SciTech Connect

    John Darrell Bess

    2008-06-01

    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

  5. Computer simulation of magnetization-controlled shunt reactors for calculating electromagnetic transients in power systems

    SciTech Connect

    Karpov, A. S.

    2013-01-15

    A computer procedure for simulating magnetization-controlled dc shunt reactors is described, which enables the electromagnetic transients in electric power systems to be calculated. It is shown that, by taking technically simple measures in the control system, one can obtain high-speed reactors sufficient for many purposes, and dispense with the use of high-power devices for compensating higher harmonic components.

  6. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere...

  7. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere...

  8. Development of a Robust Tri-Carbide Fueled Reactor for Multimegawatt Space Power and Propulsion Applications

    SciTech Connect

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

    2004-08-11

    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.

  9. Application of Molten Salt Reactor Technology to MMW In-Space NEP and Surface Power Missions

    NASA Technical Reports Server (NTRS)

    Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen (Technical Monitor)

    2002-01-01

    Anticipated manned nuclear electric propulsion (NEP) and planetary surface power missions will require multimegawatt nuclear reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional multimegawatt gas-cooled and liquid metal concepts.

  10. 76 FR 74630 - Making Changes to Emergency Plans for Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-12-01

    ... COMMISSION 10 CFR Parts 50 and 52 RIN 3150-AI10 Making Changes to Emergency Plans for Nuclear Power Reactors... Emergency Plans for Nuclear Power Reactors.'' This guide describes a method that the NRC staff considers... (DG)-1237 was published in the Federal Register on May 18, 2009 (74 FR 23220), for a 60 day...

  11. High Power Density Blanket Design Study for Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Huang, J. H.; Zhu, Y. K.; Deng, P. Zh.

    2003-06-01

    A conceptual design study of a high power density blanket has been carried out. The Fusion Experimental Breeder, FEB, is adopted as the reference reactor. The neutron wall loading is 0.5 MW/m2. The blanket is cooled by 10 MPa helium in tube. The concept of LiPb eutectic/transuranium oxide suspension is adopted. The neutronics design is performed to provide the design basis, and it gives an energy multiplication of 37 and a flattened power density distribution with a peak value of 70 W/m3. Multiple cooling panels are introduced to reduce the peak temperature of the blanket. In spite of up to 15 cooling panels, the blanket module is calculated using the ANSYS code and analytically as well. The results are consistent with each other and can meet the thermal criteria. However, structural calculation results from ANSYS did not satisfy the criterion: The blanket structure design is then improved by using curved cooling panels to model the structure in detail. Temperature distribution is obtained using the Pro/Mechanica code. Detailed structural analyses are also done by this code. Some satisfactory results are obtained.

  12. Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

    SciTech Connect

    Renier, J.A.

    2002-04-17

    Burnable poisons are used in all modern nuclear reactors to permit higher loading of fuel without the necessity of an overly large control rod system. This not only permits a longer core life but can also be used to level the power distribution. Commercial nuclear reactors commonly use B{sub 4}C in separate non-fueled rods and more recently, zirconium boride coatings on the fuel pellets or gadolinium oxide mixed with the fuel. Although the advantages are great, there are problems with using these materials. Boron, which is an effective neutron absorber, transmutes to lithium and helium upon absorption of a neutron. Helium is insoluble and is eventually released to the interior of the fuel rod, where it produces an internal pressure. When sufficiently high, this pressure stress could cause separation of the cladding from the fuel, causing overly high centerline temperatures. Gadolinium has several very strongly absorbing isotopes, but not all have large cross sections and result in residual burnable poison reactivity worth at the end of the fuel life. Even if the amount of this residual absorber is small and the penalty in operation small, the cost of this penalty, even if only several days, can be very high. The objective of this investigation was to study the performance of single isotopes in order to reduce the residual negative reactivity left over at the end of the fuel cycle. Since the behavior of burnable poisons can be strongly influenced by their configuration, four forms for the absorbers were studied: homogeneously mixed with the fuel, mixed with only the outer one-third of the fuel pellet, coated on the perimeter of the fuel pellets, and alloyed with the cladding. In addition, the numbers of fuel rods containing burnable poison were chosen as 8, 16, 64, and 104. Other configurations were chosen for a few special cases. An enrichment of 4.5 wt% {sup 235}U was chosen for most cases for study in order to achieve a 4-year fuel cycle. A standard pressurized

  13. Capillary-Pumped Passive Reactor Concept for Space Nuclear Power

    SciTech Connect

    Dr. Thomas F. Lin; Dr. Thomas G. Hughes; Christopher G. Miller

    2008-05-30

    To develop the passively-cooled space reactor concept using the capillary-induced lithium flow, since molten lithium possesses a very favorable surface tension characteristic. In space where the gravitational field is minimal, the gravity-assisted natural convection cooling is not effective nor an option for reactor heat removal, the capillary induced cooling becomes an attractive means of providing reactor cooling.

  14. Coupled Monte Carlo neutronics and thermal hydraulics for power reactors

    SciTech Connect

    Bernnat, W.; Buck, M.; Mattes, M.; Zwermann, W.; Pasichnyk, I.; Velkov, K.

    2012-07-01

    The availability of high performance computing resources enables more and more the use of detailed Monte Carlo models even for full core power reactors. The detailed structure of the core can be described by lattices, modeled by so-called repeated structures e.g. in Monte Carlo codes such as MCNP5 or MCNPX. For cores with mainly uniform material compositions, fuel and moderator temperatures, there is no problem in constructing core models. However, when the material composition and the temperatures vary strongly a huge number of different material cells must be described which complicate the input and in many cases exceed code or memory limits. The second problem arises with the preparation of corresponding temperature dependent cross sections and thermal scattering laws. Only if these problems can be solved, a realistic coupling of Monte Carlo neutronics with an appropriate thermal-hydraulics model is possible. In this paper a method for the treatment of detailed material and temperature distributions in MCNP5 is described based on user-specified internal functions which assign distinct elements of the core cells to material specifications (e.g. water density) and temperatures from a thermal-hydraulics code. The core grid itself can be described with a uniform material specification. The temperature dependency of cross sections and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. Applications will be shown for the stationary part of the Purdue PWR benchmark using ATHLET for thermal- hydraulics and for a generic Modular High Temperature reactor using THERMIX for thermal- hydraulics. (authors)

  15. 76 FR 78173 - Options for Developing the Regulatory Basis for Streamlining Non-Power Reactor License Renewal...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-12-16

    ..., Division of Policy and Rulemaking, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission... Regulatory Commission. Jessie F. Quichocho, Division of Policy and Rulemaking, Office of Nuclear Reactor...- Power Reactor License Renewal and Non-Power Reactor Emergency Preparedness AGENCY: Nuclear...

  16. Underground nuclear power plant with the VK-300 reactor as a substituting power source for Zheleznogorsk, Russia

    SciTech Connect

    Adamov, E.O.; Lebedev, V.A.; Kuznetsov, Yu.N.; Samarkin, A.A.; Tokarev, Yu.I.

    1996-07-01

    Zheleznogorsk is situated near the territorial center -- Krasnoyarsk on the Yenisei river. Mining and chemical complex is the main industrial enterprise of the town, which has been constructed for generation and used for isolation of weapons-grade plutonium. Heat supply to the chemical complex and town at the moment is largely provided by nuclear co-generation plant (NCGP) on the basis of the ADEh-2 dual-purpose reactor, generating 430 Gcal/h of heat and, partially, by coal backup peak-load boiler houses. NCGP also provides 73% of electric power consumed. In line with agreements between Russia and USA on strategic arms reduction and phasing out of weapons-grade plutonium production, decommissioning of the ADEh-2 reactor by 2000 is planned. Thus, a problem arises relative to compensation for electric and thermal power generation for the needs of the town and industrial enterprises, which is now supplied by the reactor. A nuclear power plant constructed on the same site as a substituting power source should be considered as the most practical option. Basic requirements to the reactor of substituting nuclear power plant are as follows. It is to be a new generation reactor on the basis of verified technologies, having an operating prototype optimal for underground siting and permitting utmost utilization of the available mining workings and those being disengaged. NCGP with the reactor is to be constructed in the time period required and is to become competitive with other possible power sources. Analysis has shown that the VK-300 simplified vessel-type boiling reactor meets the requirements made in the maximum extent. Its design is based on the experience of the VK-50 reactor operation for a period of 30 years in Dimitrovgrad (Russia) and allows for experience in the development of the SBWR type reactors. The design of the reactor is discussed.

  17. 77 FR 74697 - Meeting of the ACRS, Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-17

    ... published in the Federal Register on October 18, 2012, (77 FR 64146- 64147). Detailed meeting agendas and.... Evolutionary Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Evolutionary Power Reactor...

  18. Diversion assumptions for high-powered research reactors. ISPO C-50 Phase 1

    SciTech Connect

    Binford, F.T.

    1984-01-01

    This study deals with diversion assumptions for high-powered research reactors -- specifically, MTR fuel; pool- or tank-type research reactors with light-water moderator; and water, beryllium, or graphite reflectors, and which have a power level of 25 MW(t) or more. The objective is to provide assistance to the IAEA in documentation of criteria and inspection observables related to undeclared plutonium production in the reactors described above, including: criteria for undeclared plutonium production, necessary design information for implementation of these criteria, verification guidelines including neutron physics and heat transfer, and safeguards measures to facilitate the detection of undeclared plutonium production at large research reactors.

  19. Activation analysis of the PULSAR-II fusion power reactor

    SciTech Connect

    Khater, H.Y.

    1995-12-31

    The PULSAR-II pulsed tokamak power plant design utilizes a blanket made of the vanadium alloy, V-5Cr-5Ti, and cooled with liquid lithium. The shield is made of a mixture of the low activation austenitic steel (Tenelon) and vanadium. The blanket is assumed to be replaced every 5.6 full power years (FPY) and the shield is assumed to stay in place for 30 FPY. The activity induced in the blanket at the end of its lifetime is higher than the activity induced in the shield after 30 FPY. At shutdown, the blanket and shield activities are 2678 MCi and 1747 MCi, respectively. One year after shutdown the shield activity drops to 18 MCi compared to 84 MCi for the blanket. The total decay heat generated in the blanket at the end of its lifetime is 34.7 MW and drops to 17.6 MW within an hour. At shutdown, 25.3 MW of decay heat are generated in the shield, dropping to only 0.1 MW within the first year. One week after shutdown, the values of the integrated decay heat are 1770 GJ for the blanket and 469 GJ for the shield. The radwaste classification of the reactor structure is evaluated according to both the NRC 10CFR61 and Fetter waste disposal concentration limits. After 5.6 years of irradiation, the blanket will only qualify for Class C low level waste. After 30 years of operation, the shield will also qualify for disposal as Class C waste. Only remote maintenance will be allowed inside the containment building.

  20. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    SciTech Connect

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  1. Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations

    SciTech Connect

    Margaret A. Marshall; John D. Bess

    2013-02-01

    A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O2 fuelled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  2. Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor

    NASA Astrophysics Data System (ADS)

    Sager, G. T.; Wong, C. P. C.; Kapich, D. D.; McDonald, C. F.; Schleicher, R. W.

    1993-11-01

    The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankine and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The close cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed.

  3. SP-100 space reactor power system for lunar, Mars, and robotic exploration

    NASA Technical Reports Server (NTRS)

    Mondt, Jack F.

    1992-01-01

    The SP-100 power system is described which was developed for three missions, namely, Pluto Orbiter with nuclear electric propulsion; human-rated surface reactor power system for lunar and Mars exploration; and earth surveillance with an integrated nuclear electric propulsion system. The reactor power systems technology is being developed to meet these requirements so that the technical database, design tools, and specifications will be applicable to these missions. The SP-100 power system design includes the following subsystems: reactor, reactor instrumentation and control, shield, heat transport, converter, heat rejection, power conditioning control and distribution, and mechanical/structural. Particular attention is given to a demonstration mission aimed at validating technology readiness for robotic, lunar, and Mars operational missions.

  4. SP-100 space reactor power system for lunar, Mars, and robotic exploration

    NASA Astrophysics Data System (ADS)

    Mondt, Jack F.

    1992-08-01

    The SP-100 power system is described which was developed for three missions, namely, Pluto Orbiter with nuclear electric propulsion; human-rated surface reactor power system for lunar and Mars exploration; and earth surveillance with an integrated nuclear electric propulsion system. The reactor power systems technology is being developed to meet these requirements so that the technical database, design tools, and specifications will be applicable to these missions. The SP-100 power system design includes the following subsystems: reactor, reactor instrumentation and control, shield, heat transport, converter, heat rejection, power conditioning control and distribution, and mechanical/structural. Particular attention is given to a demonstration mission aimed at validating technology readiness for robotic, lunar, and Mars operational missions.

  5. Geoneutrinos and reactor antineutrinos at SNO+

    NASA Astrophysics Data System (ADS)

    Baldoncini, M.; Strati, V.; Wipperfurth, S. A.; Fiorentini, G.; Mantovani, F.; McDonough, W. F.; Ricci, B.

    2016-05-01

    In the heart of the Creighton Mine near Sudbury (Canada), the SNO+ detector is foreseen to observe almost in equal proportion electron antineutrinos produced by U and Th in the Earth and by nuclear reactors. SNO+ will be the first long baseline experiment to measure a reactor signal dominated by CANDU cores (~55% of the total reactor signal), which generally burn natural uranium. Approximately 18% of the total geoneutrino signal is generated by the U and Th present in the rocks of the Huronian Supergroup-Sudbury Basin: the 60% uncertainty on the signal produced by this lithologic unit plays a crucial role on the discrimination power on the mantle signal as well as on the geoneutrino spectral shape reconstruction, which can in principle provide a direct measurement of the Th/U ratio in the Earth.

  6. FFTF primary system transition to natural circulation from low reactor power

    SciTech Connect

    Bouchey, G.D.; Additon, S.L.; Nutt, W.T.

    1980-01-01

    Plans for reactor and primary loop natural circulation testing in the Fast Flux Test Facility (FFTF) are summarized. Detailed pretest planning with an emphasis on understanding the implications of process noise and model uncertainties for model verification and test acceptance are discussed for a transition to natural circulation in the reactor core and primary heat transport loops from initial conditions of 5% of rated reactor power and 75% of full flow.

  7. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked... 10 Energy 1 2011-01-01 2011-01-01 false Primary Reactor Containment Leakage Testing for...

  8. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked... 10 Energy 1 2013-01-01 2013-01-01 false Primary Reactor Containment Leakage Testing for...

  9. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked... 10 Energy 1 2010-01-01 2010-01-01 false Primary Reactor Containment Leakage Testing for...

  10. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked... 10 Energy 1 2014-01-01 2014-01-01 false Primary Reactor Containment Leakage Testing for...

  11. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked... 10 Energy 1 2012-01-01 2012-01-01 false Primary Reactor Containment Leakage Testing for...

  12. Space reactor/Stirling cycle systems for high power Lunar applications

    SciTech Connect

    Schmitz, P.D.; Mason, L.S.

    1994-09-01

    NASA`s Space Exploration Initiative (SEI) has proposed the use of high power nuclear power systems on the lunar surface as a necessary alternative to solar power. Because of the long lunar night ({approximately} 14 earth days) solar powered systems with the requisite energy storage in the form of regenerative fuel cells or batteries becomes prohibitively heavy at high power levels ({approximately} 100 kWe). At these high power levels nuclear power systems become an enabling technology for variety of missions. One way of producing power on the lunar surface is with an SP-100 class reactor coupled with Stirling power converters. In this study, analysis and characterization of the SP-100 class reactor coupled with Free Piston Stirling Power Conversion (FPSPC) system will be performed. Comparison of results with previous studies of other systems, particularly Brayton and Thermionic, are made.

  13. A small, 1400 deg Kelvin, reactor for Brayton space power systems

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Mayo, W.

    1972-01-01

    A preliminary cost estimate for a small reactor in Brayton space power systems with (u-233)n or (pu-239)n as the fuel in the T-111 fuel elements totaled to about four million dollars; considered is a 22.8 in. diameter reactor with 247 fuel elements.

  14. 77 FR 38338 - Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-27

    ..., which utilized a forced-circulation, direct-cycle boiling water reactor as its heat source. The plant is... March 27, 2009 (74 FR 13926). The revised regulation stated that it was applicable to all Part 50... COMMISSION Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain...

  15. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Technical specifications on effluents from nuclear power...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power...

  16. Small nuclear reactors for power and process heat

    SciTech Connect

    Olds, F.C.

    1984-11-01

    There are a number of small reactor programs around the world, with a sizable variety of designs ranging from 0.2 MWt heat-only units to 440-MWe agro-nuclear complexes. Some are in operation, some are proposed. There are light and heavy water reactors, a liquid metal cooled unit with integral hot reprocessing/refabrication, and an HTGR built with 100-MWe modules. There are several inherently safe designs. Shop fabrication is emphasized for cost and quality control, and close-in siting is practiced here and there. The feasibility of using small reactors differs markedly, depending on type of service and on country-specific factors.

  17. Heat pipe cooled reactors for multi-kilowatt space power supplies

    NASA Astrophysics Data System (ADS)

    Ranken, W. A.; Houts, M. G.

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFE's) to radiator heat pipes.

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

    NASA Technical Reports Server (NTRS)

    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.

  19. 75 FR 7634 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the Subcommittee on Power Uprates...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-22

    ... October 14, 2009 (74 FR 58268-58269). Detailed meeting agendas and meeting transcripts are available on... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the Subcommittee on Power Uprates... Safeguards. BILLING CODE 7590-01-P...

  20. Preliminary design of reactor power systems for the manned space base.

    NASA Technical Reports Server (NTRS)

    Mckhann, G. G.; Coggi, J. V.; Diamond, S. D.

    1972-01-01

    The results of design integration studies of uranium-zirconium hydride (UZr-Hx) reactor power systems for the NASA space base study program are presented. The power conversion systems investigated include the Brayton cycle, the organic Rankine cycle, the SNAP-8 mercury Rankine cycle, and thermoelectric (PbTe). The proposed space base has a 10-year life and requires 100 kWe of power. Two 50-kWe power systems with a nominal replacement life of 5 years are utilized. Parametric design data such as life, weight, radiator area, reactor outlet-temperature, reactor thermal power, and power conversion system efficiency are presented and used for the design and integration of the system with the space base.

  1. First wall and blanket design for a high wall loading compact tokamak power reactor

    SciTech Connect

    Sviatoslavsky, I.N.; Abdel-Khalik, S.I.; Corradini, M.L.; El-Afify, M.; Huh, K.Y.; Kuleinski, G.L.; Wittenberg, L.J.

    1985-07-01

    Among the specific limitations which tend to complicate a compact high wall loading (HWL) tokamak reactor design are high surface and nuclear heating, compactness leading to crowded components, unlikely breeding on the inboard side and frequent first wall/blanket replacement. This paper describes the mechanical, thermal hydraulic and tritium aspects of an improved blanket design for a high ..beta.. (20%), high wall loading (R 10 MW/m/sup 2/) compact fusion power reactor of 1000 MW /sub th/ power output.

  2. Gravity Scaling of a Power Reactor Water Shield

    NASA Technical Reports Server (NTRS)

    Reid, Robert S.; Pearson, J. Boise

    2007-01-01

    A similarity analysis on a water-based reactor shield examined the effect of gravity on free convection between a reactor shield inner and outer vessel boundaries. Two approaches established similarity between operation on the Earth and the Moon: 1) direct scaling of Rayleigh number equating gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant. Nusselt number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n).

  3. Hot zero power reactor calculations using the Insilico code

    NASA Astrophysics Data System (ADS)

    Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; Johnson, Seth R.; Pandya, Tara M.; Godfrey, Andrew T.

    2016-06-01

    In this paper we describe the reactor physics simulation capabilities of the Insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that Insilico using an SPN solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various pressurized water reactor problems. Comparison to both Monte Carlo calculations and measured plant data is provided.

  4. Advanced-power-reactor design concepts and performance characteristics

    NASA Technical Reports Server (NTRS)

    Davison, H. W.; Kirchgessner, T. A.; Springborn, R. H.; Yacobucci, H. G.

    1974-01-01

    Five reactor cooling concepts which allow continued reactor operation following a single rupture of the coolant system are presented for application with the APR. These concepts incorporate convective cooling, double containment, or heat pipes to ensure operation after a coolant line rupture. Based on an evaluation of several control system concepts, a molybdenum clad, beryllium oxide sliding reflector located outside the pressure vessel is recommended.

  5. Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system

    NASA Technical Reports Server (NTRS)

    Jefferies, K. S.; Tew, R. C.

    1974-01-01

    A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

  6. THE DEVELOPMENT OF RADIATION EMBRITTLEMENT MODELS FOR U.S. POWER REACTOR PRESSURE VESSEL STEELS

    SciTech Connect

    Wang, Jy-An John; Rao, Nageswara S

    2006-01-01

    The information fusion technique is used to develop radiation embrittlement prediction models for reactor pressure vessel (RPV) steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six parameters {Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature {are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

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

    DOEpatents

    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.

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

    DOEpatents

    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.

  9. Ballooning of CANDU pressure tubes: Hydride blisters and iodine

    SciTech Connect

    Forrest, C.F.

    1997-12-31

    Reports on a project that continues a test program to determine the effects of in-service degradation and abnormal environmental conditions on the ballooning behavior of pressure tubes in CANDU nuclear reactors. Investigators tested nine zirconium-niobium pressure tube specimens to study the effects of hydride blisters and uniformly distributed hydrogen, and the composition of the internal pressurizing gas, including argon, steam, a steam-iodine mixture, and hydrogen. Specimen internal pressure and temperature ramp rate ranged from one megapascal and one degree K per second to 9.6 megapascals and 35 K per second. Three reference tests were carried out with as-received material. Temperatures on the outside and inside surfaces of the specimens, and circumferential and longitudinal strains, were recorded during the transients. Post-test longitudinal, circumferential, and wall thickness strains were also measured.

  10. Underground collocation of nuclear power plant reactors and repository to facilitate the post-renaissance expansion of nuclear power

    SciTech Connect

    Myers, Carl W; Elkins, Ned Z

    2008-01-01

    Underground collocation of nuclear power reactors and the nuclear waste management facilities supporting those reactors, termed an underground nuclear park (UNP), appears to have several advantages compared to the conventional approach to siting reactors and waste management facilities. These advantages include the potential to lower reactor capital and operating cost, lower nuclear waste management cost, and increase margins of physical security and safety. Envirorunental impacts related to worker health, facility accidents, waste transportation, and sabotage and terrorism appear to be lower for UNPs compared to the current approach. In-place decommissioning ofUNP reactors appears to have cost, safety, envirorunental and waste disposal advantages. The UNP approach has the potential to lead to greater public acceptance for the deployment of new power reactors. Use of the UNP during the post-nuclear renaissance time frame has the potential to enable a greater expansion of U.S. nuclear power generation than might otherwise result. Technical and economic aspects of the UNP concept need more study to determine the viability of the concept.

  11. Multi-reactor configurations for multi-megawatt spacecraft power supplies

    NASA Technical Reports Server (NTRS)

    George, Jeffrey A.

    1990-01-01

    Several conceptual designs for a multimegawatt space nuclear power supply system were developed on the basis of hardware and technology from the existing multihundred kilowatt-class SP-100 reactor program. It is shown that net power outputs in the multimegawatt range can be achieved by using a modular multireactor configuration of several SP-100-derived nuclear power supplies. A variety of geometries were examined for their applicability to the multireactor configuration, showing that modular multireactor systems have the advantage of an increased redundancy in the power system, as compared with a single-reactor system, allowing higher reliabilities than those achievable with single-reactor systems. Results are presented on the Mars Cargo Mission analysis, showing that modularity allows the option of redeployment of power systems in Mars and facilitates refurbishment and turnaround of NEP transfer vehicles.

  12. Sodium coolant purification systems for a nuclear power station equipped with a BN-1200 reactor

    NASA Astrophysics Data System (ADS)

    Alekseev, V. V.; Kovalev, Yu. P.; Kalyakin, S. G.; Kozlov, F. A.; Kumaev, V. Ya.; Kondrat'ev, A. S.; Matyukhin, V. V.; Pirogov, E. P.; Sergeev, G. P.; Sorokin, A. P.; Torbenkova, I. Yu.

    2013-05-01

    Both traditional coolant purification methods (by means of traps and sorbents for removing cesium), the use of which supported successful operation of nuclear power installations equipped with fast-neutron reactors with a sodium coolant, and the possibility of removing oxygen from sodium through the use of hot traps are analyzed in substantiating the purification system for a nuclear power station equipped with a BN-1200 reactor. It is shown that a cold trap built into the reactor vessel must be a mandatory component of the reactor plant primary coolant circuit's purification system. The use of hot traps allows oxygen to be removed from the sodium coolant down to permissible concentrations when the nuclear power station operates in its rated mode. The main lines of works aimed at improving the performance characteristics of cold traps are suggested based on the results of performed investigations.

  13. Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

    SciTech Connect

    Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G.

    2012-04-04

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.

  14. Thermonuclear inverse magnetic pumping power cycle for stellarator reactor

    DOEpatents

    Ho, Darwin D.; Kulsrud, Russell M.

    1991-01-01

    The plasma column in a stellarator is compressed and expanded alternatively in minor radius. First a plasma in thermal balance is compressed adiabatically. The volume of the compressed plasma is maintained until the plasma reaches a new thermal equilibrium. The plasma is then expanded to its original volume. As a result of the way a stellarator works, the plasma pressure during compression is less than the corresponding pressure during expansion. Therefore, negative work is done on the plasma over a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils. Direct electrical energy is obtained from this voltage. Alternatively, after the compression step, the plasma can be expanded at constant pressure. The cycle can be made self-sustaining by operating a system of two stellarator reactors in tandem. Part of the energy derived from the expansion phase of a first stellarator reactor is used to compress the plasma in a second stellarator reactor.

  15. Nuclear reactor descriptions for space power systems analysis

    NASA Technical Reports Server (NTRS)

    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.

  16. Threshold self-powered gamma detector for use as a monitor of power in a nuclear reactor

    DOEpatents

    LeVert, Francis E.; Cox, Samson A.

    1978-01-01

    A self-powered gamma monitor for placement near the core of a nuclear reactor comprises a lead prism surrounded by a coaxial thin nickel sheet, the combination forming a collector. A coaxial polyethylene electron barrier encloses the collector and is separated from the nickel sheet by a vacuum region. The electron barrier is enclosed by a coaxial stainless steel emitter which, in turn, is enclosed within a lead casing. When the detector is placed in a flux of gamma rays, a measure of the current flow in an external circuit between emitter and collector provides a measure of the power level of the reactor.

  17. DynMo: Dynamic Simulation Model for Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed; Tournier, Jean-Michel

    2005-02-01

    A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK® 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.

  18. A CFD Model for High Pressure Liquid Poison Injection for CANDU-6 Shutdown System No. 2

    SciTech Connect

    Bo Wook Rhee; Chang Jun Jeong; Hye Jeong Yun; Dong Soon Jang

    2002-07-01

    In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a CFD code, to simulate the formation of the poison jet curtain inside the moderator tank. For validation, an attempt was made to validate this model against a poison injection experiment performed at BARC. As conclusion this set of models is judged to be appropriate. (authors)

  19. SVBR-100 module-type fast reactor of the IV generation for regional power industry

    NASA Astrophysics Data System (ADS)

    Zrodnikov, A. V.; Toshinsky, G. I.; Komlev, O. G.; Stepanov, V. S.; Klimov, N. N.

    2011-08-01

    In the report the following is presented: basic conceptual provisions of the innovative nuclear power technology (NPT) based on modular fast reactors (FR) SVBR-100, summarized results of calculations of the reactor, analysis of the opportunities of multi-purpose application of such reactor facilities (RF) including export potentials with due account of nonproliferation requirements. The most important features of the proposed NPT analyzed in the report are as follows: (1) integral (monoblock) arrangement of the primary circuit equipment with entire elimination of the primary circuit pipelines and valves that considerably reduces the construction and assembly works period and coupling with high boiling point of lead-bismuth coolant (LBC) deterministically eliminates accidents of the LOCA type, (2) option for 100 MWe power and dimensions of the reactor provide: on the one hand, an opportunity to transport the reactor monoblock in factory-readiness by railway as well as other kinds of transport, on the other hand, core breeding ratio (CBR) exceeds 1 while MOX-fuel is used. The preferable area of application of RF SVBR-100 is regional and small power requiring power-units of electric power in a range of (100-600) MW, which could be used for cogeneration-based district heating while locating them nearby cities as well as for generation of electric power in a mode of load tracking in the regions with low network systems.

  20. Study of UO/sub 2/ wafer fuel for very high-power research reactors

    SciTech Connect

    Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

    1980-11-01

    The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to < 20% in those research and test reactors presently using highly enriched uranium fuel. UO/sub 2/ caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO/sub 2/ wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full-power operation up to a burnup of 1.09 x 10/sup 21/ fis/cm/sup 3/; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, and wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. 12 figures, 7 tables.

  1. Space Nuclear Reactor Electric Power. (Latest citations from the Aerospace database). Published Search

    SciTech Connect

    Not Available

    1994-05-01

    The 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. (Contains 250 citations and includes a subject term index and title list.)

  2. Power Generation from Nuclear Reactors in Aerospace Applications

    NASA Technical Reports Server (NTRS)

    English, Robert E.

    1982-01-01

    Power generation in nuclear powerplants in space is addressed. In particular, the states of technology of the principal competitive concepts for power generation are assessed. The possible impact of power conditioning on power generation is also discussed. For aircraft nuclear propulsion, the suitability of various technologies is cursorily assessed for flight in the Earth's atmosphere; a program path is suggested to ease the conditions of first use of aircraft nuclear propulsion.

  3. Power generation from nuclear reactors in aerospace applications

    SciTech Connect

    English, R.E.

    1982-01-01

    Power generation in nuclear powerplants in space is addressed. In particular, the states of technology of the principal competitive concepts for power generation are assessed. The possible impact of power conditioning on power generation is also discussed. For aircraft nuclear propulsion, the suitability of various technologies is cursorily assessed for flight in the Earth's atmosphere. A program path is suggested to ease the conditions of first use of aircraft nuclear propulsion.

  4. FALCON nuclear-reactor-pumped laser program and wireless power transmission

    SciTech Connect

    Lipinski, R.J.; Pickard, P.S.

    1992-12-31

    FALCON is a high-power, reactor-pumped laser concept. The major strengths of a reactor-pumped laser are (1) simple, modular construction, (2) long-duration, closed-cycle capability, (3) self-contained power, (4) compact size, and (5) a variety of wavelengths (from visible to infrared). Reactor-pumped lasing has been demonstrated experimentally in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Powers up to 300 W for 2 ms have been demonstrated. Projected beam quality for FALCON is good enough that frequency doubling at reasonable efficiencies could be expected to yield wavelengths at 353, 363, 636, 867, 896, 1016, 1315, 1325, and 1685 nm. Appropriate missions for FALCON are described and include power beaming to satellites, the moon, and unmanned surveillance planes; lunar mapping; space debris removal; and laser propulsion.

  5. FALCON nuclear-reactor-pumped laser program and wireless power transmission

    SciTech Connect

    Lipinski, R.J.; Pickard, P.S.

    1992-01-01

    FALCON is a high-power, reactor-pumped laser concept. The major strengths of a reactor-pumped laser are (1) simple, modular construction, (2) long-duration, closed-cycle capability, (3) self-contained power, (4) compact size, and (5) a variety of wavelengths (from visible to infrared). Reactor-pumped lasing has been demonstrated experimentally in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Powers up to 300 W for 2 ms have been demonstrated. Projected beam quality for FALCON is good enough that frequency doubling at reasonable efficiencies could be expected to yield wavelengths at 353, 363, 636, 867, 896, 1016, 1315, 1325, and 1685 nm. Appropriate missions for FALCON are described and include power beaming to satellites, the moon, and unmanned surveillance planes; lunar mapping; space debris removal; and laser propulsion.

  6. Nuclear Reactors for Space Power, Understanding the Atom Series.

    ERIC Educational Resources Information Center

    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…

  7. Evaluation of Launch Accident Safety Options for Low-Power Surface Reactors

    SciTech Connect

    Fung Poon, Cindy; Poston, David I.

    2006-01-20

    Safety options for surface reactors of less than 800 kW (thermal power) are analyzed. The concepts under consideration are heat pipe cooled reactors fueled with either uranium nitride or uranium dioxide. This study investigates the impact of launch accident criteria on the system mass, while ensuring the mechanical integrity and reliability of the system through launch accident scenarios. The four criticality scenarios analyzed for shutdown determination are dry sand surround with reflectors stripped, water submersion on concrete, water submersion with all control drums in, and the nominal shutdown reactor condition. Additionally the following two operational criteria are analyzed: reactor is warm and swelled, and reactor is warm and swelled with one drum in (where swelled includes both thermal mechanical expansion and irradiation induced swelling of the fuel)

  8. UWTOR-M, a stellarator power reactor utilizing modular coils

    NASA Astrophysics Data System (ADS)

    Sviatoslavsky, I. N.; Vansciver, S. W.; Kulcinski, G. I.

    1981-10-01

    The parametric considerations which led to the UWTOR-M reference design point are described. The design has 18 twisted coils utilizing a multipolarity of 3, a major radius of 24 m, a coil radius of 4.77 m and a plasma aspect ratio of 14. An assumed (ALPHA) of 5% was used. This configuration leads to a rotational transform on the edge of 1.125 giving favorable plasma physics conditions. The natural stellarator divertor is used for impurity control in conjunction with innovative high performance divertor targets. A unique blanket design which minimizes tritium inventory in the reactor is proposed. A scheme for servicing the first wall/blanket and other reactor components is described.

  9. Fuel supply of nuclear power industry with the introduction of fast reactors

    NASA Astrophysics Data System (ADS)

    Muraviev, E. V.

    2014-12-01

    The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

  10. Hot zero power reactor calculations using the Insilico code

    DOE PAGESBeta

    Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; Johnson, Seth R.; Pandya, Tara M.; Godfrey, Andrew T.

    2016-03-18

    In this paper we describe the reactor physics simulation capabilities of the insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that the insilico SPN solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various PWR problems. Comparison to both Monte Carlo calculations and measured plant data is provided.

  11. 3-D kinetics simulations of the NRU reactor using the DONJON code

    SciTech Connect

    Leung, T. C.; Atfield, M. D.; Koclas, J.

    2006-07-01

    The NRU reactor is highly heterogeneous, heavy-water cooled and moderated, with online refuelling capability. It is licensed to operate at a maximum power of 135 MW, with a peak thermal flux of approximately 4.0 x 10{sup 18} n.m{sup -2} . s{sup -1}. In support of the safe operation of NRU, three-dimensional kinetics calculations for reactor transients have been performed using the DONJON code. The code was initially designed to perform space-time kinetics calculations for the CANDU{sup R} power reactors. This paper describes how the DONJON code can be applied to perform neutronic simulations for the analysis of reactor transients in NRU, and presents calculation results for some transients. (authors)

  12. Thermionic reactor power system: Effects of radiation on integration with Manned Space Station

    NASA Technical Reports Server (NTRS)

    Gietzen, A. J.; Heath, C. A.; Perry, L. W.

    1972-01-01

    The application of a thermionic reactor power system to the modular space station is described. The nominal net power is 40 kWe, with the power system designed to be applicable over the power range from 25 to 60 kWe. The power system is designed to be launched by the space shuttle. Radiation protection is provided by LiH neutron shielding and W gamma shielding in a shaped 4 pion configuration, i.e., the reactor is shielded on all sides but not to equal extent. Isodose contours are presented for the region around the modular space station. Levels and spectral distribution of radiation are given for later evaluation of effects on space station experiments. Parametric data on the effects of separation distance on power system mass are presented.

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

    NASA Astrophysics Data System (ADS)

    Harty, Richard B.; Durand, Richard E.; Mason, Lee S.

    1991-09-01

    An integration study was performed by coupling an SP-100 reactor to either a Brayton or Stirling power conversion subsystem. 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 standby unit. Integratiaon-design studies indicated that either the Brayton or Stirling power conversion subsystem could be integrated with the SP-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 offset the Stirling integration advantage. From a performance consideration, the Brayton had a 9-percent mass advantage and the Stirling a 50-percent radiator-area advantage.

  14. BDDR, a new CEA technological and operating reactor database

    SciTech Connect

    Soldevilla, M.; Salmons, S.; Espinosa, B.

    2013-07-01

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

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

    NASA Technical Reports Server (NTRS)

    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.

  16. Conceptual Design of a 100-kWe Space Nuclear Reactor Power System with High-Power AMTEC

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2003-01-01

    Alkali Metal Thermal-to-Electric Conversion (AMTEC), although currently at a Technology Readiness Level-3 (TRL-3), has an excellent potential for use in Space Nuclear Reactor Power (SNRP) systems for NASA's deep-space exploration missions. In addition to operating at a conversion efficiency > 20%, representing the highest fraction (> 60%) of Carnot efficiency of all other static and dynamic conversion technology options, the relatively high heat rejection radiator temperature (650-700 K) reduces the size and mass of the radiator and of the SNRP system. A high-power AMTEC unit design has been developed and optimized for operating at reactor exit temperatures <= 1180 K and radiator temperature <= 680 K. Depending on the reactor exit temperature, the nominal electrical power of the AMTEC unit, measuring 594 mm × 410 mm × 115 mm and weighting 44.3 kg, could be as high as 5.6 kWe, with a margin of >= 5% for an additional load-following increase. A conceptual design of a 100 kWe SNRP system with these high-power AMTEC units is developed and presented in this paper. The total mass of major subsystems, including the converters, nuclear reactor, shadow radiation shield, and radiator, is calculated and compared with that for the SP-100. Despite the large specific mass of the AMTEC units compared to the SiGe thermoelectrics in the SP-100 system, the lower masses of the reactor, radiation shield, and radiator make the present AMTEC-SNRP system > 26% lighter, for the same electrical power. An optimized AMTEC-SNRP system could potentially operate at a specific power > 30 We/kg (or specific mass < 33 kg/kWe), use non-refractory structures of super-steel alloys with well-know properties, relatively low density, low Ductile-To-Brittle (DTB) transition temperatures, and good compatibility with space and planetary environments containing CO2 and oxygen. The radiator area for the baseline 100 kWe AMTEC SNRP system is < 27 m2, which, together with operating the potassium heat pipes

  17. Materials technology for an advanced space power nuclear reactor concept: Program summary

    NASA Technical Reports Server (NTRS)

    Gluyas, R. E.; Watson, G. K.

    1975-01-01

    The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

  18. Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

    NASA Technical Reports Server (NTRS)

    Fox, T. A.

    1973-01-01

    An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

  19. Design Concept for a Nuclear Reactor-Powered Mars Rover

    NASA Technical Reports Server (NTRS)

    Elliott, John; Poston, Dave; Lipinski, Ron

    2007-01-01

    A report presents a design concept for an instrumented robotic vehicle (rover) to be used on a future mission of exploration of the planet Mars. The design incorporates a nuclear fission power system to provide long range, long life, and high power capabilities unachievable through the use of alternative solar or radioisotope power systems. The concept described in the report draws on previous rover designs developed for the 2009 Mars Science laboratory (MSL) mission to minimize the need for new technology developments.

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

    SciTech Connect

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

    2004-02-04

    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.

  1. The SP-100 space reactor as a power source for Mars exploration missions

    NASA Technical Reports Server (NTRS)

    Isenberg, Lon; Heller, Jack A.

    1989-01-01

    This paper argues that many of the power requirements of complex, relatively long-duration space missions such as the exploration of Mars may best be met through the use of power systems which use nuclear reactors as a thermal energy source. The development of such a power system, the SP-100, and its application in Mars mission scenarios is described. The missions addressed include a freighter mission and a mission involving exploration of the Martian surface.

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

    NASA Astrophysics Data System (ADS)

    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.

  3. Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor

    NASA Technical Reports Server (NTRS)

    Mayo, W.; Lantz, E.

    1973-01-01

    A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.

  4. Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

    NASA Astrophysics Data System (ADS)

    Hu, Jian; Jiang, Nan; Li, Jie; Shang, Kefeng; Lu, Na; Wu, Yan; Mizuno, Akira

    2016-03-01

    The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. supported by National Natural Science Foundation of China (No. 51177007), the Joint Funds of National Natural Science Foundation of China (No. U1462105), and Dalian University of Technology Fundamental Research Fund of China (No. DUT15RC(3)030)

  5. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    NASA Astrophysics Data System (ADS)

    Rohanda, Anis; Waris, Abdul

    2015-04-01

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on 16O(n,p)16N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  6. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    SciTech Connect

    Rohanda, Anis; Waris, Abdul

    2015-04-16

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  7. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

    NASA Technical Reports Server (NTRS)

    George, Jeffrey A.

    1991-01-01

    A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

  8. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

    SciTech Connect

    George, J.A.

    1991-09-01

    A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

  9. Shelding analysis for a manned Mars rover powered by an Sp-100 type reactor

    NASA Astrophysics Data System (ADS)

    Morley, Nicholas J.; El-Genk, Mohamed S.

    1991-01-01

    Shield design is one of the most crucial tasks in the integration of a nuclear reactor power system to a manned Mars rover. A multilayered W and LiH shield is found to minimize the shield mass and satisfy the dose rate limit of 30 rem/y to the rover crew. The effect on dose rate of tungsten layers thicknesses and position within the lithium hydride shields is investigated. Due to the large cross section for the W (n,γ) reaction, secondary gammas become a significant radiation source. The man-rated shield mass for the Mars rover vehicle is correlated to the reactor thermal power. The correlation fits to within 9% of the calculated shield mass and results in an uncertainty of <4% in the overall rover mass. The shield mass varied from 8600 kg to 20580 kg for a reactor thermal power of 100 to 1000 kWt, respectively.

  10. Shielding analysis for a manned Mars rover powered by an SP-100 type reactor

    NASA Astrophysics Data System (ADS)

    Morley, Nicholas J.; El-Genk, Mohamed S.

    Shield design is one of the most crucial tasks in the integration of a nuclear reactor power system to a manned Mars rover. A multilayered W and LiH shield is found to minimize the shield mass and satisfy the dose rate limit of 30 rem/y to the rover crew. The effect on dose rate of tungsten layers thicknesses and position within the lithium hydride shields is investigated. Due to the large cross section for the W (n,gamma) reaction, secondary gammas become a significant radiation source. The man-rated shield mass for the Mars rover vehicle is correlated to the reactor thermal power. The correlation fits to within 9 percent of the calculated shield mass and results in an uncertainty of below 4 percent in the overall rover mass. The shield mass varied from 8600 kg to 20580 kg for a reactor thermal power of 100 to 1000 kW(t), respectively.

  11. Shielding analysis for a manned Mars rover powered by an SP-100 type reactor

    NASA Technical Reports Server (NTRS)

    Morley, Nicholas J.; El-Genk, Mohamed S.

    1991-01-01

    Shield design is one of the most crucial tasks in the integration of a nuclear reactor power system to a manned Mars rover. A multilayered W and LiH shield is found to minimize the shield mass and satisfy the dose rate limit of 30 rem/y to the rover crew. The effect on dose rate of tungsten layers thicknesses and position within the lithium hydride shields is investigated. Due to the large cross section for the W (n,gamma) reaction, secondary gammas become a significant radiation source. The man-rated shield mass for the Mars rover vehicle is correlated to the reactor thermal power. The correlation fits to within 9 percent of the calculated shield mass and results in an uncertainty of below 4 percent in the overall rover mass. The shield mass varied from 8600 kg to 20580 kg for a reactor thermal power of 100 to 1000 kW(t), respectively.

  12. Shield materials recommended for space power nuclear reactors

    NASA Technical Reports Server (NTRS)

    Kaszubinski, L. J.

    1973-01-01

    Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

  13. Linear signal-compensated amplifier for reactor power measuring channels

    SciTech Connect

    Khaleeq, M. Tahir; Atique-ur-Rahman,; Ahmed, Eijaz

    2006-07-15

    A linear amplifier with automatic signal compensation has been developed for nuclear channels. The amplifier controls its sensitivity automatically according to the reference input within the desired settings and has automatic signal compensation capability for use in the nuclear channels. The amplifier will be used in the existing safety channel of Pakistan Research Reactor-1, where the system has an independent sensitivity control unit for manual compensation of xenon effect. The new amplifier will improve the safety of the system. The amplifier is tested and the results found are in very good agreement with the designed specifications. This article presents design and construction of the amplifier and test results.

  14. Linear signal-compensated amplifier for reactor power measuring channels

    NASA Astrophysics Data System (ADS)

    Khaleeq, M. Tahir; Atique-ur-Rahman, Ahmed, Eijaz

    2006-07-01

    A linear amplifier with automatic signal compensation has been developed for nuclear channels. The amplifier controls its sensitivity automatically according to the reference input within the desired settings and has automatic signal compensation capability for use in the nuclear channels. The amplifier will be used in the existing safety channel of Pakistan Research Reactor-1, where the system has an independent sensitivity control unit for manual compensation of xenon effect. The new amplifier will improve the safety of the system. The amplifier is tested and the results found are in very good agreement with the designed specifications. This article presents design and construction of the amplifier and test results.

  15. High Efficiency Nuclear Power Plants using Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITS of 950 K and 1200 K are presented. Power plant performance data were obtained for TITS ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo -generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  16. High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  17. AECL/US INERI - Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Power Reactors -- Fuel Requirements and Down-Select Report

    SciTech Connect

    William Carmack; Randy D. Lee; Pavel Medvedev; Mitch Meyer; Michael Todosow; Holly B. Hamilton; Juan Nino; Simon Philpot; James Tulenko

    2005-06-01

    potential advantage for more efficient destruction of plutonium and minor actinides (MA) relative to MOX fuel. Greater efficiency in plutonium reduction results in greater flexibility in managing plutonium inventories and in developing strategies for disposition of MA, as well as a potential for fuel cycle cost savings. Because fabrication of plutonium-bearing (and MA-bearing) fuel is expensive relative to UO{sub 2} in terms of both capital and production, cost benefit can be realized through a reduction in the number of plutonium-bearing elements required for a given burn rate. In addition, the choice of matrix material may be manipulated either to facilitate fuel recycling or to make plutonium recovery extremely difficult. In addition to plutonium/actinide management, an inert matrix fuel having high thermal conductivity may have operational and safety benefits; lower fuel temperatures could be used to increase operating and safety margins, uprate reactor power, or a combination of both. The CANDU reactor offers flexibility in plutonium management and MA burning by virtue of online refueling, a simple bundle design, and good neutron economy. A full core of inert matrix fuel containing either plutonium or a plutonium-actinide mix can be utilized, with plutonium destruction efficiencies greater than 90%, and high (>60%) actinide destruction efficiencies. The Advanced CANDU Reactor (ACR) could allow additional possibilities in the design of an IMF bundle, since the tighter lattice pitch and light-water coolant reduce or eliminate the need to suppress coolant void reactivity, allowing the center region of the bundle to include additional fissile material and to improve actinide burning. The ACR would provide flexibility for management of plutonium and MA from the existing LWR fleet, and would be complementary to the AFCI program in the U.S. Many of the fundamental principles concerning the use of IMF are nearly identical in LWRs and the ACR, including fuel

  18. Expert system driven fuzzy control application to power reactors

    SciTech Connect

    Tsoukalas, L.H.; Berkan, R.C.; Upadhyaya, B.R.; Uhrig, R.E.

    1990-12-31

    For the purpose of nonlinear control and uncertainty/imprecision handling, fuzzy controllers have recently reached acclaim and increasing commercial application. The fuzzy control algorithms often require a ``supervisory`` routine that provides necessary heuristics for interface, adaptation, mode selection and other implementation issues. Performance characteristics of an on-line fuzzy controller depend strictly on the ability of such supervisory routines to manipulate the fuzzy control algorithm and enhance its control capabilities. This paper describes an expert system driven fuzzy control design application to nuclear reactor control, for the automated start-up control of the Experimental Breeder Reactor-II. The methodology is verified through computer simulations using a valid nonlinear model. The necessary heuristic decisions are identified that are vitally important for the implemention of fuzzy control in the actual plant. An expert system structure incorporating the necessary supervisory routines is discussed. The discussion also includes the possibility of synthesizing the fuzzy, exact and combined reasoning to include both inexact concepts, uncertainty and fuzziness, within the same environment.

  19. Expert system driven fuzzy control application to power reactors

    SciTech Connect

    Tsoukalas, L.H.; Berkan, R.C.; Upadhyaya, B.R.; Uhrig, R.E.

    1990-01-01

    For the purpose of nonlinear control and uncertainty/imprecision handling, fuzzy controllers have recently reached acclaim and increasing commercial application. The fuzzy control algorithms often require a supervisory'' routine that provides necessary heuristics for interface, adaptation, mode selection and other implementation issues. Performance characteristics of an on-line fuzzy controller depend strictly on the ability of such supervisory routines to manipulate the fuzzy control algorithm and enhance its control capabilities. This paper describes an expert system driven fuzzy control design application to nuclear reactor control, for the automated start-up control of the Experimental Breeder Reactor-II. The methodology is verified through computer simulations using a valid nonlinear model. The necessary heuristic decisions are identified that are vitally important for the implemention of fuzzy control in the actual plant. An expert system structure incorporating the necessary supervisory routines is discussed. The discussion also includes the possibility of synthesizing the fuzzy, exact and combined reasoning to include both inexact concepts, uncertainty and fuzziness, within the same environment.

  20. Next Generation CANDU: Conceptual Design for a Short Construction Schedule

    SciTech Connect

    Hopwood, Jerry M.; Love, Ian J.W.; Elgohary, Medhat; Fairclough, Neville

    2002-07-01

    Atomic Energy of Canada Ltd. (AECL) has very successful experience in implementing new construction methods at the Qinshan (Phase III) twin unit CANDU 6 plant in China. This paper examines the construction method that must be implemented during the conceptual design phase of a project if short construction schedules are to be met. A project schedule of 48 months has been developed for the nth unit of NG (Next Generation) CANDU with a 42 month construction period from 1. Concrete to In-Service. An overall construction strategy has been developed involving paralleling project activities that are normally conducted in series. Many parts of the plant will be fabricated as modules and be installed using heavy lift cranes. The Reactor Building (RB), being on the critical path, has been the focus of considerable assessment, looking at alternative ways of applying the construction strategy to this building. A construction method has been chosen which will result in excess of 80% of internal work being completed as modules or as very streamlined traditional construction. This method is being further evaluated as the detailed layout proceeds. Other areas of the plant have been integrated into the schedule and new construction methods are being applied to these so that further modularization and even greater paralleling of activities will be achieved. It is concluded that the optimized construction method is a requirement, which must be implemented through all phases of design to make a 42 month construction schedule a reality. If the construction methods are appropriately chosen, the schedule reductions achieved will make nuclear more competitive. (authors)

  1. Rotating-bed reactor as a power source for EM gun applications

    SciTech Connect

    Powell, J.; Botts, T.; Stickley, C.M.; Meth, S.

    1980-01-01

    Electromagnetic gun applications of the Rotating Bed Reactor (RBR) are examined. The RBR is a compact (approx. 1 m/sup 3/), (up to several thousand MW(th)), high-power reactor concept, capable of producing a high-temperature (up to approx. 300/sup 0/K) gas stream with a MHD generator coupled to it, the RBR can generate electric power (up to approx. 1000 MW(e)) in the pulsed or cw modes. Three EM gun applications are investigated: a rail gun thruster for orbit transfer, a rapid-fire EM gun for point defense, and a direct ground-to-space launch. The RBR appears suitable for all applications.

  2. Lunar in-core thermionic nuclear reactor power system conceptual design

    NASA Technical Reports Server (NTRS)

    Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.

    1991-01-01

    This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Exploration Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

  3. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    SciTech Connect

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  4. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... Commission will not issue a construction permit after March 27, 1986 for a non-power reactor where the... proposed reactor will have a unique purpose as defined in § 50.2. (2) Unless the Commission has determined... reactor has a unique purpose, each licensee authorized to possess and use HEU fuel in connection with...

  5. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... Commission will not issue a construction permit after March 27, 1986 for a non-power reactor where the... proposed reactor will have a unique purpose as defined in § 50.2. (2) Unless the Commission has determined... reactor has a unique purpose, each licensee authorized to possess and use HEU fuel in connection with...

  6. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... Commission will not issue a construction permit after March 27, 1986 for a non-power reactor where the... proposed reactor will have a unique purpose as defined in § 50.2. (2) Unless the Commission has determined... reactor has a unique purpose, each licensee authorized to possess and use HEU fuel in connection with...

  7. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... Commission will not issue a construction permit after March 27, 1986 for a non-power reactor where the... proposed reactor will have a unique purpose as defined in § 50.2. (2) Unless the Commission has determined... reactor has a unique purpose, each licensee authorized to possess and use HEU fuel in connection with...

  8. Triga Mark III Reactor Operating Power and Neutron Flux Study by Nuclear Track Methodology

    NASA Astrophysics Data System (ADS)

    Espinosa, G.; Golzarri, J. I.; Raya-Arredondo, R.; Cruz-Galindo, S.; Sajo-Bohus, L.

    The operating power of a TRIGA Mark III reactor was studied using Nuclear Track Methodology (NTM). The facility has a Highly Enriched Uranium core that provides a neutron flux of around 2 x 1012 n cm-2 s-1 in the TO-2 irradiation channel. The detectors consisted of a Landauer® CR-39 (allyl diglycol polycarbonate) chip covered with a 3 mm Plexiglas® converter. After irradiation, the detectors were chemically etched in a 6.25M-KOH solution at 60±1 °C for 6 h. Track density was determined by a custom-made Digital Image Analysis System. The results show a direct proportionality between reactor power and average nuclear track density for powers in the range 0.1-7 kW. Data reproducibility and relatively low uncertainty (±3%) were achieved. NTM is a simple, fast and reliable technique that can serve as a complementary procedure to measure reactor operating power. It offers the possibility of calibrating the neutron flux density in any low power reactor.

  9. Loop system with gas control of the power production in the MR reactor

    SciTech Connect

    Andreev, V.I.; Kolyadin, V.I.; Smirnov, A.I.; Yakovlev, V.V.

    1987-01-01

    An unsolved problem in reactor design is premature fuel-pin failure on account of mechanical interaction between the fuel and the sheath under nonstationary operating conditions. To examine the effects of this interaction on the viability, the authors have built an experimental system with the MR reactor. To provide for varying the power production over wide ranges, gas regulation based on /sup 3/He as neutron absorber is used. A map of core loading in the MR reactor is provided and variation in power in the experimental fuel assembly in accordance with /sup 3/He pressure and control location is shown. A structural diagram shows the reactor apparatus with gas power control in the experimental pin assembly. The relative changes in channel power in relation to neutron absorber pressure in GCU in channel 1-4 are presented. The results are offered on the power variation in the experimental assembly and reactivity as functions of /sup 3/He pressure in the GCU, together with the calculated data.

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

    NASA Technical Reports Server (NTRS)

    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

  11. Novel, Integrated Reactor / Power Conversion System (LMR-AMTEC)

    SciTech Connect

    Pablo Rubiolo, Principal Investigator

    2003-03-21

    The main features of this project were the development of a long life (up to 10 years) Liquid Metal Reactor (LMR) and a static conversion subsystem comprising an Alkali Metal Thermal-to-Electric (AMTEC) topping cycle and a ThermoElectric (TE) Bottom cycle. Various coupling options of the LMR with the energy conversion subsystem were explored and, base in the performances found in this analysis, an Indirect Coupling (IC) between the LMR and the AMTEC/TE converters with Alkali Metal Boilers (AMB) was chosen as the reference design. The performance model of the fully integrated sodium-and potassium-AMTEC/TE converters shows that a combined conversion efficiency in excess of 30% could be achieved by the plant. (B204)

  12. Partial site release at a power reactor facility.

    PubMed

    Darman, Joseph; Whitney, Michael; Dubiel, Richard

    2004-01-01

    U.S. NRC licensed facilities undergoing decommissioning may wish to remove portions of their site from the jurisdiction of their license, prior to final license termination. The method of partial site release, relevant to radiological conditions, described herein employs NUREG-1505 methodology for demonstrating indistinguishability from background. The partial site release process was also informed by NRC Regulatory Issue Summary 2000-19 "Partial Release of Reactor Site for Unrestricted Use Before NRC Approval of the License Termination Plan." However, the focus of this discussion is the radiological aspects of partial site release, relevant to the implementation of NUREG-1505 methodology for demonstrating indistinguishability from background, based on the 137Cs concentrations at the site and a suitable background reference area. This type of approach was found acceptable by the NRC, and the partial site release was granted. PMID:14695009

  13. Measurement of SRS reactor recirculation pump performance using pump motor power

    SciTech Connect

    Whitehouse, J.C.

    1994-03-01

    In order to accurately predict reactor hydraulic behavior during a hypothetical Loss-of-Coolant-Accident (LOCA) the performance of reactor coolant pumps under off-design conditions must be understood. The LOCA of primary interest for the Savannah River Site (SRS) production reactors involves the aspiration of air into the recirculated heavy water flow as reactor tank inventory is lost (system temperatures are too low to result in significant flashing of water coolant into steam). Entrained air causes degradation in the performance of the large recirculation pumps. The amount of degradation is a parameter used in computer codes which predict the course of the accident. This paper describes the analysis of data obtained during in-reactor simulated LOCA tests, and presents the head degradation curve for the SRS reactor recirculation pumps. The greatest challenge of the analysis was to determine a reasonable estimate of mixture density at the pump suction. Specially designed three-beam densitometers were used to determine mixture density. Since it was not feasible to place them in the most advantageous location the measured pump motor power, along with other techniques (pressure corrected gamma densitometer void fraction), were used to calculate the average mixture density at the pump impeller. These techniques provided good estimates of pump suction mixture density. Measurements from more conventional instruments were used to arrive at the value of pump two-component head over a wide range of flows. The results were significantly different from previous work with commercial reactor recirculation pumps.

  14. SUSEE: A Compact, Lightweight Space Nuclear Power System Using Present Water Reactor Technology

    NASA Astrophysics Data System (ADS)

    Maise, George; Powell, James; Paniagua, John

    2006-01-01

    The SUSEE space reactor system uses existing nuclear fuels and the standard steam cycle to generate electrical and thermal power for a wide range of in-space and surface applications, including manned bases, sub-surface mobile probes to explore thick ice deposits on Mars and the Jovian moons, and mobile rovers. SUSEE cycle efficiency, thermal to electric, ranges from ~20 to 24%, depending on operating parameters. Rejection of waste heat is by a lightweight condensing radiator that can be launched as a compact rolled-up package and deployed into flat panels when appropriate. The 50 centimeter diameter SUSEE reactor can provide power over the range of 10 kW(e) to 1 MW(e) for a period of 10 years. Higher power outputs are possible using slightly larger reactors. System specific weight (reactor, turbine, generator, piping, and radiator is ~3 kg/kW(e). Two SUSEE reactor options are described, based on the existing Zr/O2 cermet and the UH3/ZrH2 TRIGA nuclear fuels.

  15. SUSEE: A Compact, Lightweight Space Nuclear Power System Using Present Water Reactor Technology

    SciTech Connect

    Maise, George; Powell, James; Paniagua, John

    2006-01-20

    The SUSEE space reactor system uses existing nuclear fuels and the standard steam cycle to generate electrical and thermal power for a wide range of in-space and surface applications, including manned bases, sub-surface mobile probes to explore thick ice deposits on Mars and the Jovian moons, and mobile rovers. SUSEE cycle efficiency, thermal to electric, ranges from {approx}20 to 24%, depending on operating parameters. Rejection of waste heat is by a lightweight condensing radiator that can be launched as a compact rolled-up package and deployed into flat panels when appropriate. The 50 centimeter diameter SUSEE reactor can provide power over the range of 10 kW(e) to 1 MW(e) for a period of 10 years. Higher power outputs are possible using slightly larger reactors. System specific weight (reactor, turbine, generator, piping, and radiator) is {approx}3 kg/kW(e). Two SUSEE reactor options are described, based on the existing Zr/O2 cermet and the UH3/ZrH2 TRIGA nuclear fuels.

  16. The Development of Radiation Embrittlement Models for U. S. Power Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John; Rao, Nageswara S; Konduri, Savanthi

    2007-01-01

    A new approach of utilizing information fusion technique is developed to predict the radiation embrittlement of reactor pressure vessel steels. The Charpy transition temperature shift data contained in the Power Reactor Embrittlement Database is used in this study. Six parameters {Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature {are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

  17. Study of reactor Brayton power systems for nuclear electric spacecraft

    NASA Technical Reports Server (NTRS)

    1979-01-01

    The feasibility of using Brayton power systems for nuclear electric spacecraft was investigated. The primary performance parameters of systems mass and radiator area were determined for systems from 100 to 1000 kW sub e. Mathematical models of all system components were used to determine masses and volumes. Two completely independent systems provide propulsion power so that no single-point failure can jeopardize a mission. The waste heat radiators utilize armored heat pipes to limit meteorite puncture. The armor thickness was statistically determined to achieve the required probability of survival. A 400 kW sub e reference system received primary attention as required by the contract. The components of this system were defined and a conceptual layout was developed with encouraging results. An arrangement with redundant Brayton power systems having a 1500 K (2240 F) turbine inlet temperature was shown to be compatible with the dimensions of the space shuttle orbiter payload bay.

  18. Preparations for shifting the power units of nuclear power stations equipped with RBMK-1000 reactors for operation with a 2 year interval between repairs

    NASA Astrophysics Data System (ADS)

    Yanchenko, Yu. A.; Filimontsev, Yu. N.; Osipova, S. E.; Dement'ev, V. N.; Butorin, S. L.; Petrov, A. A.

    2010-05-01

    A general approach for carrying out works on justifying the shifting of power units used at nuclear power stations equipped with RBMK-1000 reactors for operation with an increased interval between repairs is formulated. The technical and organizational measures ensuring reliable operation of equipment and pipelines and acceptable safety of power units at nuclear power stations equipped with RBMK-1000 reactors in the new schedule of operation are described.

  19. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    SciTech Connect

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

  20. Advanced Power Conversion Efficiency in Inventive Plasma for Hybrid Toroidal Reactor

    NASA Astrophysics Data System (ADS)

    Hançerlioğullari, Aybaba; Cini, Mesut; Güdal, Murat

    2013-08-01

    Apex hybrid reactor has a good potential to utilize uranium and thorium fuels in the future. This toroidal reactor is a type of system that facilitates the occurrence of the nuclear fusion and fission events together. The most important feature of hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity good power transformation productivity, the magnitude of the reactor's operational duration, low failure percentage, short maintenance time and the inclusion of the system's simple technology and material. The analysis has been made using the MCNP Monte Carlo code and ENDF/B-V-VI nuclear data. Around the fusion chamber, molten salts Flibe (LI2BeF4), lead-lithium (PbLi), Li-Sn, thin-lityum (Li20Sn80) have used as cooling materials. APEX reactor has modeled in the torus form by adding nuclear materials of low significance in the specified percentages between 0 and 12 % to the molten salts. In this study, the neutronic performance of the APEX fusion reactor using various molten salts has been investigated. The nuclear parameters of Apex reactor has been searched for Flibe (LI2BeF4) and Li-Sn, for blanket layers. In case of usage of the Flibe (LI2BeF4), PbLi, and thin-lityum (Li20Sn80) salt solutions at APEX toroidal reactors, fissile material production per source neutron, tritium production speed, total fission rate, energy reproduction factor has been calculated, the results obtained for both salt solutions are compared.

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

    SciTech Connect

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

    2006-01-20

    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, respect0011ive.

  2. Requirements for a common nuclear propulsion and power reactor for human exploration missions to Mars

    NASA Astrophysics Data System (ADS)

    Cataldo, Robert L.; Borowski, Stanley K.

    1998-01-01

    Requirements for propulsion and power systems capable of achieving a safe, reliable, robust and affordable human Mars exploration mission have been identified. Nuclear systems have been identified that can meet the challenges of short trip times, reduced number of launch vehicles, potential for ``all propulsive'' maneuvers, abundant in-space power and low mass, volume and deployed area, and energy rich surface power. Reduced total systems cost will also be mandatory to achieve affordable human exploration of Mars. Hence, it is desirable to design a space propulsion and surface power reactor with the greatest degree of commonality as possible with the goal of reducing total system costs.

  3. Requirements for a common nuclear propulsion and power reactor for human exploration missions to Mars

    SciTech Connect

    Cataldo, Robert L.; Borowski, Stanley K.

    1998-01-15

    Requirements for propulsion and power systems capable of achieving a safe, reliable, robust and affordable human Mars exploration mission have been identified. Nuclear systems have been identified that can meet the challenges of short trip times, reduced number of launch vehicles, potential for 'all propulsive' maneuvers, abundant in-space power and low mass, volume and deployed area, and energy rich surface power. Reduced total systems cost will also be mandatory to achieve affordable human exploration of Mars. Hence, it is desirable to design a space propulsion and surface power reactor with the greatest degree of commonality as possible with the goal of reducing total system costs.

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

    NASA Technical Reports Server (NTRS)

    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.

  5. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    SciTech Connect

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-30

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  6. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-01

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  7. Manned mars rover powered by a nuclear reactor; Radiation shield analysis

    SciTech Connect

    Morley, N.J.; El-Genk, M. . Dept. of Chemical and Nuclear Engineering)

    1992-08-01

    This paper discusses a key element in the conceptual design of a nuclear reactor power system for a manned Mars rover is the analysis, design, and integration of the radiation shield. A shield analysis is carried out to characterize the thickness and spacing of shield layers to provide the minimum mass configuration that meets a dose rate requirement of 300 mSv/yr. The analysis utilizes a two-dimensional transport code to model the reactor and to provide a source term that is subsequently used to calculate dose rates as a function of reactor power level and shield layer thickness. Results show that a multilayered tungsten and lithium hydride (LiH) shield would satisfy the dose rate limit of 300 mSv/yr (30 rem/yr) to the rover crew. The position of two tungsten and LiH layers is varied to minimize secondary gamma-ray production and to optimize shield mass.

  8. Estimates of the financial consequences of nuclear-power-reactor accidents

    SciTech Connect

    Strip, D.R.

    1982-09-01

    This report develops preliminary techniques for estimating the financial consequences of potential nuclear power reactor accidents. Offsite cost estimates are based on CRAC2 calculations. Costs are assigned to health effects as well as property damage. Onsite costs are estimated for worker health effects, replacement power, and cleanup costs. Several classes of costs are not included, such as indirect costs, socio-economic costs, and health care costs. Present value discounting is explained and then used to calculate the life cycle cost of the risks of potential reactor accidents. Results of the financial consequence estimates for 156 reactor-site combinations are summarized, and detailed estimates are provided in an appendix. The results indicate that, in general, onsite costs dominate the consequences of potential accidents.

  9. Insights from Investigations of In-Vessel Retention for High Powered Reactors

    SciTech Connect

    Joy L. Rempe

    2005-10-01

    In a three-year U.S. - Korean International Nuclear Energy Research Initiative (INERI), state-of-the-art analytical tools and key U.S. and Korean experimental facilities were used to explore two options, enhanced ERVC performance and the use of internal core catchers, that have the potential to increase the margin for in-vessel retention (IVR) in high power reactors (up to 1500 MWe). This increased margin has the potential to improve plant economics (owing to reduced regulatory requirements) and increase public acceptance (owing to reduced plant risk). Although this program focused upon the Korean Advanced Power Reactor -- 1400 MWe (APR 1400) design, recommentations were developed so that they can easily be applied to a wide range of existing and advanced reactor designs. This paper summarizes new data gained for evaluating the margin associated with various options investigated in this program. Insights from analyses completed with this data are also highlighted.

  10. A modular gas-cooled cermet reactor system for planetary base power

    SciTech Connect

    Jahshan, S.N.; Borkowski, J.A. )

    1993-01-15

    Fission nuclear power is foreseen as the source for electricity in planetary colonization and exploration. A six module gas-cooled, cermet-fueled reactor is proposed that can meet the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers six modular Brayton cycles that compare favorably with the SP-100-based Brayton cycle.

  11. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.58 Safety/security...

  12. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Applications for Licenses, Certifications, and Regulatory Approvals; Form;...

  13. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... environmental conditions created by the burning of hydrogen. Environmental conditions caused by...

  14. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Inspections, Records, Reports, Notifications § 50.72 Immediate notification requirements for operating nuclear...

  15. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Applications for Licenses, Certifications, and Regulatory Approvals; Form; Contents; Ineligibility of Certain Applicants...

  16. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.58 Safety/security...

  17. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.58 Safety/security...

  18. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Applications for Licenses, Certifications, and Regulatory Approvals; Form; Contents; Ineligibility of Certain Applicants...

  19. Thermal-hydraulics and safety analysis of sectored compact reactor for lunar surface power

    SciTech Connect

    Schriener, T. M.; El-Genk, M. S.

    2012-07-01

    The liquid NaK-cooled, fast-neutron spectrum, Sectored Compact Reactor (SCoRe-N 5) concept has been developed at the Univ. of New Mexico for lunar surface power applications. It is loaded with highly enriched UN fuel pins in a triangular lattice, and nominally operates at exit and inlet coolant temperatures of 850 K and 900 K. This long-life reactor generates up to 1 MWth continuously for {>=} 20 years. To avoid a single point failure in reactor cooling, the core is divided into 6 sectors that are neutronically and thermally coupled, but hydraulically independent. This paper performs a 3-D the thermal-hydraulic analysis of SCoRe--N 5 at nominal operation temperatures and a power level of 1 MWth. In addition, the paper investigates the potential of continuing reactor operation at a lower power in the unlikely event that one sector in the core experiences a loss of coolant (LOC). Redesigning the core with a contiguous steel matrix enhances the cooling of the sector experiencing a LOC. Results show that with a core sector experiencing a LOC, SCORE-N 5 could continue operating safely at a reduced power of 166.6 kWth. (authors)

  20. Introduction to Nuclear Fusion Power and the Design of Fusion Reactors. An Issue-Oriented Module.

    ERIC Educational Resources Information Center

    Fillo, J. A.

    This three-part module focuses on the principles of nuclear fusion and on the likely nature and components of a controlled-fusion power reactor. The physical conditions for a net energy release from fusion and two approaches (magnetic and inertial confinement) which are being developed to achieve this goal are described. Safety issues associated…

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

    NASA Technical Reports Server (NTRS)

    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.

  2. Gaseous-fuel nuclear reactor research for multimegawatt power in space

    NASA Technical Reports Server (NTRS)

    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.

  3. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Applications for Licenses, Certifications, and Regulatory Approvals; Form; Contents; Ineligibility of Certain Applicants...

  4. Requalification of SPERT (Special Power Excursion Reactor Test) pins for use in university reactors

    SciTech Connect

    Snelgrove, J.L.; Domagala, R.F.; Dates, L.R.

    1986-12-01

    A series of nondestructive and destructive examinations have been performed on a representative sample of stainless steel-clad UO/sub 2/ fuel pins procured in the early-to-mid 1960s for the SPERT program. These examinations were undertaken in order to requalify the SPERT pins for use in converting university research reactors from the use of highly enriched uranium to the use of low-enriched uranium. The requalification program included visual and dimensional inspections of fuel pins and fuel pellets, radiographic inspections of welds, fill gas analyses, and chemical and spectrographic analyses of fuel and cladding materials. In general all attributes tested were within or very close to specified values, although some weld defects not covered by the original specifications were found. 1 ref., 4 figs., 11 tabs.

  5. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage. 73.55 Section 73.55 Energy NUCLEAR... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this...

  6. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage. 73.55 Section 73.55 Energy NUCLEAR... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this...

  7. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... nuclear power reactors licensed under 10 CFR parts 50 or 52 and authorized to use special nuclear...

  8. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... nuclear power reactors licensed under 10 CFR parts 50 or 52 and authorized to use special nuclear...

  9. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... nuclear power reactors licensed under 10 CFR parts 50 or 52 and authorized to use special nuclear...

  10. A neutron tomography facility at a low power research reactor

    NASA Astrophysics Data System (ADS)

    Koerner, S.; Schillinger, B.; Vontobel, P.; Rauch, H.

    2001-09-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at this beam position is 1.3×10 5 neutrons/cm 2 s and the beam diameter is 8 cm. For a 3D tomographic reconstruction of the sample interior, transmission images of the object taken from different view angles are required. Therefore, a rotary table driven by a step motor connected to a computerized motion control system has been installed at the sample position. In parallel a suitable electronic imaging device based on a neutron sensitive scintillator screen and a CCD-camera has been designed. It can be controlled by a computer in order to synchronize the software of the detector and of the rotary table with the aim of an automation of measurements. Reasonable exposure times can get as low as 20 s per image. This means that a complete tomography of a sample can be performed within one working day. Calculation of the 3D voxel array is made by using the filtered backprojection algorithm.

  11. Optimum Reflector Configurations for Minimizing Fission Power Peaking in a Lithium-Cooled, Liquid-Metal Reactor with Sliding Reflectors

    SciTech Connect

    Fensin, Michael L.; Poston, David I.

    2005-02-06

    Many design constraints limit the development of a space fission power system optimized for fuel performance, system reliability, and mission cost. These design constraints include fuel mass provisions to meet cycle-length requirements, fuel centerline and clad temperatures, and clad creep from fission gas generation. Decreasing the fission power peaking of the reactor system enhances all of the mentioned parameters. This design study identifies the cause, determines the reflector configurations for reactor criticality, and generates worth curves for minimized fission-power-peaking configuration in a lithium-cooled liquid-metal reactor that uses sliding reflectors. Because of the characteristics of the core axial power distribution and axial power distortions inherent to the sliding reflector design, minimizing the power peaking of the reactor involves placing the reflectors in a position that least distorts the axial power distribution. The views expressed in this document are those of the author and do not necessarily reflect agreement by the Government.

  12. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    SciTech Connect

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  13. Power Systems Development Facility: Performance and development of components in the transport reactor train

    SciTech Connect

    Powell, C.A.; Vimalchand, P.; Leonard, R.F.

    1998-12-31

    The Power Systems Development Facility (PSDF) will develop and demonstrate advanced power generation technologies and system components needed to improve process reliability. This paper will provide an introduction to the PSDF and discuss in detail the operation and performance of the M.W. Kellogg Company`s (MWK) Transport reactor train system components. There will also be brief discussions on the operation and performance of the Transport reactor and the Particulate Collection Device (PCD). Discussions will focus on the major operational challenges faced during the commissioning and operation of various components and the significant equipment modifications that were made to improve the reliability and performance. These include: modifications to the pulverizers, corrective actions taken to the transport air and recycle gas systems, improvements to the process gas analysis system, and changes to the steam generation package. Also included are operational findings of the particle disengagement and collection system, experiences with solids handling systems, and continued development of the reactor`s startup burner, pressure letdown valve, process air systems and impacts of corrosion downstream of the PCD. Much can be inferred from the experiences gained at the PSDF as to the impact each component or system had on the successful operation of the MWK Transport reactor train and similar technologies in the future.

  14. Space Molten Salt Reactor Concept for Nuclear Electric Propulsion and Surface Power

    NASA Astrophysics Data System (ADS)

    Eades, M.; Flanders, J.; McMurray, N.; Denning, R.; Sun, X.; Windl, W.; Blue, T.

    Students at The Ohio State University working under the NASA Steckler Grant sought to investigate how molten salt reactors with fissile material dissolved in a liquid fuel medium can be applied to space applications. Molten salt reactors of this kind, built for non-space applications, have demonstrated high power densities, high temperature operation without pressurization, high fuel burn up and other characteristics that are ideal for space fission systems. However, little research has been published on the application of molten salt reactor technology to space fission systems. This paper presents a conceptual design of the Space Molten Salt Reactor (SMSR), which utilizes molten salt reactor technology for Nuclear Electric Propulsion (NEP) and surface power at the 100 kWe to 15 MWe level. Central to the SMSR design is a liquid mixture of LiF, BeF2 and highly enriched U235F4 that acts as both fuel and core coolant. In brief, some of the positive characteristics of the SMSR are compact size, simplified core design, high fuel burn up percentages, proliferation resistant features, passive safety mechanisms, a considerable body of previous research, and the possibility for flexible mission architecture.

  15. Application of H[infinity] control theory to power control of a nonlinear reactor model

    SciTech Connect

    Suzuki, Katsuo; Shimazaki, Junya; Shinohara, Yoshikuni . Dept. of Reactor Engineering)

    1993-10-01

    The H[infinity] control theory is applied to the compensator design of a nonlinear nuclear reactor model, and the results are compared with standard linear quadratic Gaussian (LQG) control. The reactor model is assumed to be provided with a control rod drive system having the compensation of rod position feedback. The nonlinearity of the reactor model exerts a great influence on the stability of the control system, and hence, it is desirable for a power control system of a nuclear reactor to achieve robust stability and to improve the sensitivity of the feedback control system. A computer simulation based on a power control system synthesized by LQG control was performed revealing that the control system has some stationary offset and less stability. Therefore, here, attention is given to the development of a methodology for robust control that can withstand exogenous disturbances and nonlinearity in view of system parameter changes. The developed methodology adopts H[infinity] control theory in the feedback system and shows interesting features of robustness. The results of the computer simulation indicate that the feedback control system constructed by the developed H[infinity] compensator possesses sufficient robustness of control on the stability and disturbance attenuation, which are essential for the safe operation of a nuclear reactor.

  16. Neutronic design studies for an unattended, low power reactor

    SciTech Connect

    Palmer, R.G.; Durkee, J.W. Jr.

    1986-01-01

    The Los Alamos National Laboratory is involved in the design and demonstrations of a small, long-lived nuclear heat and electric power source for potential applications at remote sites where alternate fossil energy systems would not be cost effective. This paper describes the neutronic design analysis that was performed to arrive at two conceptual designs, one using thermoelectric conversion, the other using an organic Rankine cycle. To meet the design objectives and constraints a number of scoping and optimization studies were carried out. The results of calculations of control worths, temperature coefficients of reactivity and fuel depletion effects are reported.

  17. Analysis of a Helium Brayton Power Cycle for a Direct-Drive Inertial Fusion Energy Power Reactor

    NASA Astrophysics Data System (ADS)

    Wagner, Scott; Gentile, Charles; Parsells, Robert; Priniski, Craig

    2008-11-01

    Presented is a thermodynamic model analysis and optimization of a helium Brayton power cycle for direct-drive inertial fusion energy (IFE) reactor. Preliminary reactor design goals include production of 2GW of thermal power and an estimated 700MW of electricity using a tertiary indirect helium Brayton cycle. A thermodynamic analysis of the proposed helium Brayton cycle is performed using baseline technology specifications and generalized thermodynamic assumptions. Analytic equations are developed using first and second law analysis. The model constraints are the turbine inlet temperature and pressure set by the reactor temperature of ˜700^oC and current turbine specifications of 7MPa, respectively. Optimization of this model is then performed using iterative numerical programming for key variables. Previous analysis shows a 51% cycle efficiency using current technology; best estimates of near-term technology increase the cycle efficiency to 64%. Results will be presented. R. Schleicher, A. R. Raffray, C. P. Wong, ``An Assessment of the Brayton Cycle for High Performance Power Plant,'' Fusion Technology, 39 (2), 823-827, March 2001.

  18. Data bases for rapid response to power reactor problems

    SciTech Connect

    Maskewitz, B.F.

    1980-01-01

    The urgency of the TMI-2 incident demanded prompt answers to an imperious situation. In responding to these challenging circumstances, both government and industry recognized deficiencies in both availability of essential retrievable data and calculational capabilities designed to respond immediately to actual abnormal events. Each responded by initiating new programs to provide a remedy for the deficiencies and to generally improve all safety measures in the nuclear power industry. Many data bases and information centers offer generic data and other technology resources which are generally useful in support of nuclear safety programs. A few centers can offer rapid access to calculational methods and associated data and more will make an effort to do so. As a beneficial spin-off from the lessons learned from TMI-2, more technical effort and financial resources will be devoted to the prevention of accidents, and to improvement of safety measures in the immediate future and for long term R and D programs by both government and the nuclear power industry.

  19. Design of a personnel TLD badge for a power reactor beta/gamma spectrum

    SciTech Connect

    Quinn, D.M.; Labenski, T. )

    1983-10-01

    This paper reports that three basic challenges are inherent in the design of a thermoluminescent dosimeter for a power reactor beta/gamma spectrum: the dosimeter must meet the current standard for performance in laboratory testing, the dosimeter must properly respond to a power reactor spectrum that is different from that specified in the standard, and the dosimeter must function under field conditions. These challenges were met at the Indian Point 3 Nuclear Power Station by modifying the case of a commercial multi-element TLD to include varying thicknesses of tissue equivalent plastic absorbers over the elements. An algorithm was developed to correct the TLD responses for laboratory testing: however, in field use, shallow and deep dose are read directly from the TLD without the use of an algorithm.

  20. Experimental power density distribution benchmark in the TRIGA Mark II reactor

    SciTech Connect

    Snoj, L.; Stancar, Z.; Radulovic, V.; Podvratnik, M.; Zerovnik, G.; Trkov, A.; Barbot, L.; Domergue, C.; Destouches, C.

    2012-07-01

    In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the few available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)

  1. 14C content in vegetation in the vicinities of Brazilian nuclear power reactors.

    PubMed

    Dias, Cíntia Melazo; Santos, Roberto Ventura; Stenström, Kristina; Nícoli, Iêda Gomes; Skog, Göran; da Silveira Corrêa, Rosangela

    2008-07-01

    (14)C specific activities were measured in grass samples collected around Brazilian nuclear power reactors. The specific activity values varied between 227 and 299 Bq/kg C. Except for two samples which showed (14)C specific activities 22% above background values, half of the samples showed background specific activities, and the other half had a (14)C excess of 1-18%. The highest specific activities were found close to the nuclear power plants and along the main wind directions (NE and NNE). The activity values were found to decrease with increasing distance from the reactors. The unexpectedly high (14)C excess values found in two samples were related to the local topography, which favors (14)C accumulation and limits the dispersion of the plume. The results indicate a clear (14)C anthropogenic signal within 5 km around the nuclear power plants which is most prominent along northeastwards, the prevailing wind direction. PMID:18308434

  2. Utilization of the Philippine Research Reactor as a training facility for nuclear power plant operators

    SciTech Connect

    Palabrica, R.J.

    1981-01-01

    The Philippines has a 1-MW swimming-pool reactor facility operated by the Philippine Atomic Energy Commission (PAEC). The reactor is light-water moderated and cooled, graphite reflected, and fueled with 90% enriched uranium. Since it became critical in 1963 it has been utilized for research, radioisotope production, and training. It was used initially in the training of PAEC personnel and other research institutions and universities. During the last few years, however, it has played a key role in training personnel for the Philippine Nuclear Power Project (PNPP).

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

    SciTech Connect

    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

  4. A gas-cooled cermet reactor system for planetary base power

    SciTech Connect

    Jahshan, S.N.; Borkowski, J.A.

    1992-01-01

    Fission nuclear power is foreseen as the source for electricity in colonization exploration. A gas-cooled, cermet-fueled reactor is proposed that can meet many of the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers a Brayton cycle that compares well with the SP-100-based Brayton cycle. The power cycle can be upgraded further under certain siting-related conditions by the addition of a low temperature Rankine cycle.

  5. A gas-cooled cermet reactor system for planetary base power

    SciTech Connect

    Jahshan, S.N.; Borkowski, J.A.

    1992-08-01

    Fission nuclear power is foreseen as the source for electricity in colonization exploration. A gas-cooled, cermet-fueled reactor is proposed that can meet many of the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers a Brayton cycle that compares well with the SP-100-based Brayton cycle. The power cycle can be upgraded further under certain siting-related conditions by the addition of a low temperature Rankine cycle.

  6. Testing of Passive Safety System Performance for Higher Power Advanced Reactors

    SciTech Connect

    brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright

    2004-12-31

    This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

  7. Accident sequence analysis for a BWR (Boiling Water Reactor) during low power and shutdown operations

    SciTech Connect

    Whitehead, D.W.; Hake, T.M.

    1990-01-01

    Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs.

  8. Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems. M.S. Thesis

    NASA Technical Reports Server (NTRS)

    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. Neutronic feasibility of dual mode operation and smaller reactor sizes than those previously investigated are shown to be possible. A heat transfer analysis of one such reactor shows that the dual-mode concept is applicable when power generation mode thermal power levels are within the same order of magnitude as direct thrust mode thermal power levels.

  9. Applicability of 100kWe-class of space reactor power systems to NASA manned space station missions. Final Report

    SciTech Connect

    Silverman, S.W.; Willenberg, H.J.; Robertson, C.

    1985-08-01

    An assessment is made of a manned space station operating with sufficiently high power demands to require a multihundred kilowatt range electrical power system. The nuclear reactor is a competitor for supplying this power level. Load levels were selected at 150kWe and 300kWe. Interactions among the reactor electrical power system, the manned space station, the space transportation system, and the mission were evaluated. The reactor shield and the conversion equipment were assumed to be in different positions with respect to the station; on board, tethered, and on a free flyer platform. Mission analyses showed that the free flyer concept resulted in unacceptable costs and technical problems. The tethered reactor providing power to an electrolyzer for regenerative fuel cells on the space station, results in a minimum weight shield and can be designed to release the reactor power section so that it moves to a high altitude orbit where the decay period is at least 300 years. Placing the reactor on the station, on a structural boom is an attractive design, but heavier than the long tethered reactor design because of the shield weight for manned activity near the reactor.

  10. Applicability of 100kWe-class of space reactor power systems to NASA manned space station missions

    NASA Technical Reports Server (NTRS)

    Silverman, S. W.; Willenberg, H. J.; Robertson, C.

    1985-01-01

    An assessment is made of a manned space station operating with sufficiently high power demands to require a multihundred kilowatt range electrical power system. The nuclear reactor is a competitor for supplying this power level. Load levels were selected at 150kWe and 300kWe. Interactions among the reactor electrical power system, the manned space station, the space transportation system, and the mission were evaluated. The reactor shield and the conversion equipment were assumed to be in different positions with respect to the station; on board, tethered, and on a free flyer platform. Mission analyses showed that the free flyer concept resulted in unacceptable costs and technical problems. The tethered reactor providing power to an electrolyzer for regenerative fuel cells on the space station, results in a minimum weight shield and can be designed to release the reactor power section so that it moves to a high altitude orbit where the decay period is at least 300 years. Placing the reactor on the station, on a structural boom is an attractive design, but heavier than the long tethered reactor design because of the shield weight for manned activity near the reactor.

  11. Power beaming to space using a nuclear reactor-pumped laser

    SciTech Connect

    Lipinski, R.J.; Monroe, D.K.; Pickard, P.S.

    1993-10-01

    The present political and environmental climate may slow the inevitable direct utilization of nuclear power in space. In the meantime, there is another approach for using nuclear energy for space power. That approach is to let nuclear energy generate a laser beam in a ground-based nuclear reactor-pumped laser (RPL), and then beam the optical energy into space. Potential space applications for a ground-based RPL include (1) illuminating geosynchronous communication satellites in the earth`s shadow to extend their lives, (2) beaming power to orbital transfer vehicles, (3) providing power (from earth) to a lunar base during the long lunar night, and (4) removing space debris. FALCON is a high-power, steady-state, nuclear reactor-pumped laser (RPL) concept that is being developed by the Department of Energy with Sandia National Laboratories as the lead laboratory. The FALCON program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at wavelengths of 0.585, 0.703, 0.725, 1.271, 1.733, 1.792, 2.032, 2.63, 2.65, and 3.37 {mu}m with intrinsic efficiency as high as 2.5%. Frequency-doubling the 1.733{minus}{mu}m line would yield a good match for photovoltaic arrays at 0.867 {mu}m. Preliminary designs of an RPL suitable for power beaming have been completed. The MWclass laser is fairly simple in construction, self-powered, closed-cycle (no exhaust gases), and modular. This paper describes the FALCON program accomplishments and power-beaming applications.

  12. Advanced Fusion Reactors for Space Propulsion and Power Systems

    SciTech Connect

    Chapman, John J.

    2011-06-15

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles' exhaust momentum can be used directly to produce high Isp thrust and also offer possibility of power conversion into electricity. p-11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  13. Advanced Fusion Reactors for Space Propulsion and Power Systems

    NASA Technical Reports Server (NTRS)

    Chapman, John J.

    2011-01-01

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  14. TFTR (Tokamak Fusion Test Reactor) neutral beam injected power measurement

    SciTech Connect

    Kamperschroer, J.H.; Grisham, L.R.; Dudek, L.E.; Gammel, G.M.; Johnson, G.A.; Kugel, H.W.; Lagin, L.; O'Connor, T.E.; Shah, P.A.; Sichta, P.

    1989-05-01

    Energy flow within TFTR neutral beamlines is measured with a waterfall calorimetry system capable of simultaneously measuring the energy deposited within four heating beamlines (three ion sources each), or of measuring the energy deposited in a separate neutral beam test stand. Of the energy extracted from the ion source in the well instrumented test stand, 99.5 +- 3.5% can be accounted for. When the ion deflection magnet is energized, however, 6.5% of the extracted energy is lost. This loss is attributed to a spray of devious particles onto unmonitored surfaces. A 30% discrepancy is also observed between energy measurements on the internal beamline calorimeter and energy measurements on a calorimeter located in the test stand target chamber. Particle reflection from the flat plate calorimeter in the target chamber, which the incident beam strikes at a near-grazing angle of 12/degree/, is the primary loss of this energy. A slight improvement in energy accountability is observed as the beam pulse length is increased. This improvement is attributed to systematic error in the sensitivity of the energy measurement to small fluctuations on the supply water temperature. An overall accuracy of 15% is estimated for the total power injected into TFTR. Contributions to this error are uncertainties in the beam neutralization efficiency, reionization and beam scrape-off in the drift duct, and fluctuations in the temperature of the supply water. 28 refs., 9 figs., 1 tab.

  15. ANALYSIS OF A HIGH TEMPERATURE GAS-COOLED REACTOR POWERED HIGH TEMPERATURE ELECTROLYSIS HYDROGEN PLANT

    SciTech Connect

    M. G. McKellar; E. A. Harvego; A. M. Gandrik

    2010-11-01

    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  16. Apparatus and method for closed-loop control of reactor power in minimum time

    DOEpatents

    Bernard, Jr., John A.

    1988-11-01

    Closed-loop control law for altering the power level of nuclear reactors in a safe manner and without overshoot and in minimum time. Apparatus is provided for moving a fast-acting control element such as a control rod or a control drum for altering the nuclear reactor power level. A computer computes at short time intervals either the function: .rho.=(.beta.-.rho.).omega.-.lambda..sub.e '.rho.-.SIGMA..beta..sub.i (.lambda..sub.i -.lambda..sub.e ')+l* .omega.+l* [.omega..sup.2 +.lambda..sub.e '.omega.] or the function: .rho.=(.beta.-.rho.).omega.-.lambda..sub.e .rho.-(.lambda..sub.e /.lambda..sub.e)(.beta.-.rho.)+l* .omega.+l* [.omega..sup.2 +.lambda..sub.e .omega.-(.lambda..sub.e /.lambda..sub.e).omega.] These functions each specify the rate of change of reactivity that is necessary to achieve a specified rate of change of reactor power. The direction and speed of motion of the control element is altered so as to provide the rate of reactivity change calculated using either or both of these functions thereby resulting in the attainment of a new power level without overshoot and in minimum time. These functions are computed at intervals of approximately 0.01-1.0 seconds depending on the specific application.

  17. An evaluation of the ecological consequences of partial-power operation of the K Reactor, SRS

    SciTech Connect

    Gladden, J.B.; Mackey, H.E.; Paller, M.H.; Specht, W.L.; Wike, L.D.; Wilde, E.W.

    1991-06-01

    The K Reactor at the Savannah River Site (SRS) shut-down in spring 1988 for maintenance and safety upgrades. Since that time the receiving stream for thermal effluent, Indian Grave Branch and Pen Branch, have undergone a pattern of post-thermal recovery that is typical of other SRS streams following removal of thermal stress. Divesity of fish and aquatic macroinvertebrate communities has increased and available habitats have been colonized by numerous species of herbaceous and woody plants. K Reactor is scheduled to resume operation in 1991 and operate through 1992 without a cooling tower to cool the discharge. It is likely that the reactor will operate at approximately one-third to one-half of full power (800--1200 MW thermal) during this period and effluent temperatures will be substantially lower than earlier operation at full power. Monthly average discharge temperatures at half-power operation will range from approximately 42{degrees}C in winter to 49{degrees}C in summer. The volume of water discharged will not be affected by altered power levels and will average approximately 10--11 m{sup 3}/s. The ecological consequences of this mode of operation on the Indian Grave/Pen Branch stream system have been evaluated.

  18. On the fusion triple product and fusion power gain of tokamak pilot plants and reactors

    NASA Astrophysics Data System (ADS)

    Costley, A. E.

    2016-06-01

    The energy confinement time of tokamak plasmas scales positively with plasma size and so it is generally expected that the fusion triple product, nTτ E, will also increase with size, and this has been part of the motivation for building devices of increasing size including ITER. Here n, T, and τ E are the ion density, ion temperature and energy confinement time respectively. However, tokamak plasmas are subject to operational limits and two important limits are a density limit and a beta limit. We show that when these limits are taken into account, nTτ E becomes almost independent of size; rather it depends mainly on the fusion power, P fus. In consequence, the fusion power gain, Q fus, a parameter closely linked to nTτ E is also independent of size. Hence, P fus and Q fus, two parameters of critical importance in reactor design, are actually tightly coupled. Further, we find that nTτ E is inversely dependent on the normalised beta, β N; an unexpected result that tends to favour lower power reactors. Our findings imply that the minimum power to achieve fusion reactor conditions is driven mainly by physics considerations, especially energy confinement, while the minimum device size is driven by technology and engineering considerations. Through dedicated R&D and parallel developments in other fields, the technology and engineering aspects are evolving in a direction to make smaller devices feasible.

  19. SiC Semiconductor Detector Power Monitors for Space Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Reisi Fard, Mehdi; Blue, Thomas E.; Miller, Don W.

    2004-02-01

    As a part of a Department of Energy-Nuclear Engineering Research Initiative (NERI) Project, we are investigating SiC semiconductor detectors as power monitors for Generation IV power reactors. SiC detectors are well-suited as power monitors for reactors for space nuclear propulsion, due to their characteristics of small size, mass, and power consumption; mechanical ruggedness; radiation hardness; capability for high temperature operation; and potential for pulse mode operation at high count rates, which may allow for a reduction in the complexity of the reactor instrumentation and control system, as well as allow for verification of detector sensitivity, verification of channel operability, and channel self-repair. In this paper, a mathematical model of a SiC detector is presented. The model includes a description of the formation of electron-hole pairs in a SiC diode detector, using the computer code TRIM. The TRIM results are used as input to a MATLAB simulation of detector current output pulse formation, the results of which are intended for use as the input to a model of the detector channel as a whole.

  20. Vital area identification for U.S. Nuclear Regulatory Commission nuclear power reactor licensees and new reactor applicants.

    SciTech Connect

    Whitehead, Donnie Wayne; Varnado, G. Bruce

    2008-09-01

    U.S. Nuclear Regulatory Commission nuclear power plant licensees and new reactor applicants are required to provide protection of their plants against radiological sabotage, including the placement of vital equipment in vital areas. This document describes a systematic process for the identification of the minimum set of areas that must be designated as vital areas in order to ensure that all radiological sabotage scenarios are prevented. Vital area identification involves the use of logic models to systematically identify all of the malicious acts or combinations of malicious acts that could lead to radiological sabotage. The models available in the plant probabilistic risk assessment and other safety analyses provide a great deal of the information and basic model structure needed for the sabotage logic model. Once the sabotage logic model is developed, the events (or malicious acts) in the model are replaced with the areas in which the events can be accomplished. This sabotage area logic model is then analyzed to identify the target sets (combinations of areas the adversary must visit to cause radiological sabotage) and the candidate vital area sets (combinations of areas that must be protected against adversary access to prevent radiological sabotage). Any one of the candidate vital area sets can be selected for protection. Appropriate selection criteria will allow the licensee or new reactor applicant to minimize the impacts of vital area protection measures on plant safety, cost, operations, or other factors of concern.

  1. A comparative risk assessment for the Russian V213 power reactor

    SciTech Connect

    Marshall, T.D.; Hockenbury, R.W.; Honey, J.A.; Cadwallader, L.C.

    1996-04-01

    Probabilistic risk assessment methodology is applied to generate an evaluation of the relative likelihood of safe recovery following selected pressurized water reactor (PWR) design basis accidents for a Russian V213 nuclear power reactor. US-designed PWRs similar to the V213 are used for reference and comparison. This V213 risk assessment is based on comparison analyses of the following aspects: accident progression event tree success paths for typical PWR accident initiating events, safety aspects in reactor design, and perceived performance of reactor safety systems. The four initiating events considered here are: loss of offsite power with station blackout, large-break loss-of-coolant accident (LOCA), medium-break LOCA, and small-break LOCA. The success probabilities for the V213 reaching a non-core-damage state after the onset of the selected initiating events are calculated for two scenarios: (a) using actual component reliability data from US PWRs and (b) assuming common component reliability data. US PWR component reliability data are used based of the unavailability of such data for the V213 at the time of the analyses. While the use of US PWR data in this risk assessment of the V213 does strongly infer V213 comparability to US plants, the risk assessment using common component reliability does not have such a stringent limitation and is thus a separate scoping assessment of the V213 engineered safety systems. The results of the analyses suggest that the V213 has certain design features that significantly improve the reactor`s safety margin for the selected initiating events and that the V213 design has a relative risk of core damage for selected initiating events that is at least comparable to US PWRs. It is important to realize that these analyses are of a scoping nature and may be significantly influenced by important risk factors such as V213 operator training, quality control, and maintenance procedures.

  2. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    NASA Astrophysics Data System (ADS)

    Fic, Adam; Składzień, Jan; Gabriel, Michał

    2015-03-01

    Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

  3. The Information Fusion Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John; Rao, Nageswara S; Konduri, Savanthi

    2007-01-01

    The complex nonlinear dependencies observed in typical reactor pressure vessel (RPV) material embrittlement data, as well as the inherent large uncertainties and scatter in the radiation embrittlement data, make prediction of radiation embrittlement a difficult task. Conventional statistical and deterministic approaches have only resulted in rather large uncertainties, in part because they do not fully exploit domain-specific mechanisms. The domain models built by researchers in the field, on the other hand, do not fully exploit the statistical and information content of the data. As evidenced in previous studies, it is unlikely that a single method, whether statistical, nonlinear, or domain model, will outperform all others. More generally, considering the complexity of the embrittlement prediction problem, it is highly unlikely that a single best method exists and is tractable, even in theory. In this paper, we propose to combine a number of complementary methods including domain models, neural networks, and nearest neighbor regressions (NNRs). Such a combination of methods has become possible because of recent developments in measurement-based optimal fusers in the area of information fusion. The information fusion technique is used to develop radiation embrittlement prediction models for reactor RPV steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six Cu, Ni, P, neutron fluence, irradiation time, and irradiation-parameters are used in the embrittlement prediction models. The results-temperature indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The

  4. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  5. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  6. The NERVA Derivative Reactor - A multi-application space power source

    SciTech Connect

    Pierce, B.L.; Wett, J.F.; Chi, J.W.H.

    1987-01-01

    The U.S. Air Force, SDI, and NASA have identified increasing needs for electric power for all types of space missions. For many of these, only nuclear-electric can provide the lowest life cycle cost. Among the many different types of nuclear space power systems proposed, the NERVA Derivative Reactor, based on the proven NERVA/ROVER technology stands out as the most attractive. It can be integrated with closed and open cycle turbo-generators and open cycle MHD generators to provide the wide range of diverse power requirements that include multikilowatts to megawatts of steady state, baseload power and multi-megawatts of burst power for weapon systems. The NDR technology can be applied to these systems with relatively little additional engineering developments, which are primarily related to demonstrating compliance with the space nuclear safety requirements.

  7. Analysis of closed cycle megawatt class space power systems with nuclear reactor heat sources

    NASA Technical Reports Server (NTRS)

    Juhasz, A. J.; Jones, B. I.

    1987-01-01

    The analysis and integration studies of multimegawatt nuclear power conversion systems for potential SDI applications is presented. A study is summarized which considered 3 separate types of power conversion systems for steady state power generation with a duty requirement of 1 yr at full power. The systems considered are based on the following conversion cycles: direct and indirect Brayton gas turbine, direct and indirect liquid metal Rankine, and in core thermionic. A complete mass analysis was performed for each system at power levels ranging from 1 to 25 MWe for both heat pipe and liquid droplet radiator options. In the modeling of common subsystems, reactor and shield calculations were based on multiparameter correlation and an in-house analysis for the heat rejection and other subsystems.

  8. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    NASA Technical Reports Server (NTRS)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2010-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  9. An RF-Powered Micro-Reactor for Efficient Extraction and Hydrolysis

    NASA Astrophysics Data System (ADS)

    Scott, V.

    2014-12-01

    An RF sample-processing micro-reactor that was developed as part of potential in situ Exploration Missions to inner- and outer-planetary bodies was designed to utilize aqueous solutions subjected to 60 GHz radiation at 730 mW of input power to extract target organic compounds and molecular and inorganic ions as well as to hydrolyze complex polymeric materials. Successful identification and characterization of these molecules relies on the sample-processing techniques utilized alongside state-of-the-art detection and analysis. For mass and power restrictions put on space exploration missions, smaller and more efficient instruments are highly desirable. The RF micro-reactor potentially offers a simplified alternative to the typical gold-standard extractions that often use solvents, chemicals, and conditions that can vary wildly and depend on the targeted molecules. Instead, this instrument uses a single solvent ­— water — that can be "tuned" under the different experimental conditions, leveraging the operating principles of the Sub-Critical Water Extractor. Proof-of-concept experiments examining the hydrolysis of glycosidic and peptide bonds were successful in demonstrating the RF micro-reactor's capabilities. Progress toward coupling the reactor with a micro-scale sample-handling system enabling slurry delivery has been made and preliminary results on heterogeneous reactions and extractions will be presented.

  10. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    NASA Technical Reports Server (NTRS)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2009-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

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

    NASA Astrophysics Data System (ADS)

    Harty, Richard B.; Durand, Richard 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.

  12. Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

    NASA Technical Reports Server (NTRS)

    Godfroy, Thomas; Dickens, Ricky; Houts, Michael; Pearson, Boise; Webster, Kenny; Gibson, Marc; Qualls, Lou; Poston, Dave; Werner, Jim; Radel, Ross

    2011-01-01

    The Nuclear Systems Team at NASA Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and Mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program, which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter for tests at MSFC. When tested at NASA Glenn Research Center (GRC) the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumentation (temperature, pressure, flow) for data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

  13. Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

    NASA Astrophysics Data System (ADS)

    Godfroy, T.; Dickens, R.; Houts, M.; Pearson, B.; Webster, K.; Gibson, M.; Qualls, L.; Poston, D.; Werner, J.; Radel, R.

    The Nuclear Systems Team at Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter when being tested at MSFC. When tested at GRC the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumenta- tion (temperature, pressure, flow) data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

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

    SciTech Connect

    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.

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

    NASA Technical Reports Server (NTRS)

    Harty, Richard B.; Durand, Richard E.

    1993-01-01

    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.

  16. Influence of FRAPCON-1 evaluation models on fuel behavior calculations for commercial power reactors. [PWR; BWR

    SciTech Connect

    Chambers, R.; Laats, E.T.

    1981-01-01

    A preliminary set of nine evaluation models (EMs) was added to the FRAPCON-1 computer code, which is used to calculate fuel rod behavior in a nuclear reactor during steady-state operation. The intent was to provide an audit code to be used in the United States Nuclear Regulatory Commission (NRC) licensing activities when calculations of conservative fuel rod temperatures are required. The EMs place conservatisms on the calculation of rod temperature by modifying the calculation of rod power history, fuel and cladding behavior models, and materials properties correlations. Three of the nine EMs provide either input or model specifications, or set the reference temperature for stored energy calculations. The remaining six EMs were intended to add thermal conservatism through model changes. To determine the relative influence of these six EMs upon fuel behavior calculations for commercial power reactors, a sensitivity study was conducted. That study is the subject of this paper.

  17. 100-kWe lunar/Mars surface power utilizing the SP-100 reactor with dynamic conversion

    NASA Astrophysics Data System (ADS)

    Harty, Richard B.; Mason, Lee S.

    Results are presented from a study of the coupling of an SP-100 nuclear reactor with either a Stirling or Brayton power system, at the 100 kWe level, for a power generating system suitable for operation in the lunar and Martian surface environments. In the lunar environment, the reactor and primary coolant loop would be contained in a guard vessel to protect from a loss of primary loop containment. For Mars, all refractory components, including the reactor, coolant, and power conversion components will be contained in a vacuum vessel for protection against the CO2 environment.

  18. CANDU MOX initiative: Report on a stakeholders` debate

    SciTech Connect

    Gizewski, P.

    1997-12-31

    The safe, secure disposition of excess plutonium from dismantled Russian and US nuclear warheads is a significant international priority. One option being considered involves the fabrication outside of Canada of mixed oxide (MOX) fuel bundles for CANDU reactors. These bundles would contain up to 3% plutonium in oxide form, mixed with uranium oxide. This option is the subject of growing controversy, both in terms of its substance as well as the process by which the MOX option proposal has thus far unfolded. This report summarizes a meeting held to debate the MOX initiative and its implications. Participants include representatives from Atomic Energy of Canada, Ontario Hydro, federal and provincial governments, non-governmental organizations, and interested citizens. The report highlights the main features of the initiative, the nature of the arguments advanced in favor and against it, and the manner in which the debate was conducted. Issues discussed include international security implications, alternatives to the MOX scheme, MOX fuel transportation and security, health-related concerns, the regulatory process, community perspectives, and the policy process.

  19. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    SciTech Connect

    Sheryl Morton; Carl Baily; Tom Hill; Jim Werner

    2006-02-01

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a lowtemperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  20. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    SciTech Connect

    Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.; Werner, James E.

    2006-01-20

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  1. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    NASA Astrophysics Data System (ADS)

    Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.; Werner, James E.

    2006-01-01

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  2. Power Distribution Analysis for the ORNL High Flux Isotope Reactor Critical Experiment 3

    SciTech Connect

    Chandler, David; Primm, Trent; Maldonado, G Ivan

    2010-01-01

    The mission of the Reduced Enrichment for Research and Test Reactors Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low-enriched uranium (LEU) fuel and targets. Oak Ridge National Laboratory (ORNL) is currently reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction offlux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. Power distributions were calculated in the fuel elements of the HFIR, a research reactor at ORNL, via MCNP and were compared to experimentally obtained data. This study was performed to validate Monte Carlo methods for power density calculations and to observe biases. A current three-dimensional MCNP model was modified to replicate the 1965 HFIR Critical Experiment 3 (HFIRCE-3). In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. 'Foils' (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation, and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil's activity to the activity of a normalizing foil. This analysis consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the reference foil

  3. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A. ); Blake, J.E.; Rush, G.C. )

    1990-01-01

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  4. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A.; Blake, J.E.; Rush, G.C.

    1990-12-31

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  5. Experimental Evaluation of a Water Shield for a Surface Power Reactor

    NASA Technical Reports Server (NTRS)

    Pearson, J. Boise; Reid, Robert S.

    2006-01-01

    As part of the Vision for Space Exploration the end of the next decade will bring man back to the surface of the moon. One of the most critical issues for the establishment of human presence on the moon will be the availability of compact power sources. The establishment of man on the moon will require power from greater than 10's of kWt's in follow on years. Nuclear reactors are extremely we11 suited to meet the needs for power generation on the lunar or Martian surface. reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), Boron Carbide, and others. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to remove the potential for radiation streaming paths. The water shield concept relies on predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. MSFC has developed the experience and fac necessary to do this evaluation in the Early Flight Fission - Test Facility (EFF-TF).

  6. Preliminary Evaluation of Convective Heat Transfer in a Water Shield for a Surface Power Reactor

    NASA Technical Reports Server (NTRS)

    Pearson J. Boise; Reid, Robert S.

    2007-01-01

    As part of the Vision for Space Exploration, the end of the next decade will bring man back to the surface of the moon. A crucial issue for the establishment of human presence on the moon will be the availability of compact power sources. This presence could require greater than 10's of kWt's in follow on years. Nuclear reactors are well suited to meet the needs for power generation on the lunar or Martian surface. Radiation shielding is a key component of any surface power reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), and boron carbide. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to fix the location of any vapor that could form radiation streaming paths. The water shield concept relies on the predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. NASA Marshall Space Flight Center has developed the experience and facilities necessary to do this evaluation in its Early Flight Fission - Test Facility (EFF-TF).

  7. Transmutation behaviour of Eurofer under irradiation in the IFMIF test facility and fusion power reactors

    NASA Astrophysics Data System (ADS)

    Fischer, U.; Simakov, S. P.; Wilson, P. P. H.

    2004-08-01

    The transmutation behaviour of the low activation steel Eurofer was analysed for irradiation simulations in the high flux test module (HFTM) of the International Fusion Material Irradiation Facility (IFMIF) neutron source and the first wall of a typical fusion power reactor (FPR) employing helium cooled lithium lead (HCLL) and pebble bed (HCPB) blankets. The transmutation calculations were conducted with the analytical and laplacian adaptive radioactivity analysis (ALARA) code and IEAF-2001 data for the IFMIF and the EASY-2003 system for the fusion power reactor (FPR) irradiations. The analyses showed that the transmutation of the main constituents of Eurofer, including iron and chromium, is not significant. Minor constituents such as Ti, V and Mn increase by 5-15% per irradiation year in the FPR and by 10-35% in the IFMIF HFTM. Other minor constituents such as B, Ta, and W show a different transmutation behaviour resulting in different elemental compositions of the Eurofer steel after high fluence irradiations in IFMIF and fusion power reactors.

  8. Discrepancies between film and thermoluminescent dosimetry (TLD) readings at an operating power reactor

    SciTech Connect

    Quinn, D.M.

    1980-01-01

    The results of exposure measurements using film badges and thermoluminescent dosimetry (TLD) were compared at an operating nuclear power reactor. The film badge overresponded to the high-energy Nitrogen-16 gamma rays produced under power, while the TLD did not. Discussions of charged-particle equilibrium and energy dependence are included. The cause of the overresponse was determined to be the excess pair production electrons created because of the high atomic number in the lead energy-compensating shield surrounding the film and in the film itself.

  9. High Temperature Water Heat Pipes Radiator for a Brayton Space Reactor Power System

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2006-01-01

    A high temperature water heat pipes radiator design is developed for a space power system with a sectored gas-cooled reactor and three Closed Brayton Cycle (CBC) engines, for avoidance of single point failures in reactor cooling and energy conversion and rejection. The CBC engines operate at turbine inlet and exit temperatures of 1144 K and 952 K. They have a net efficiency of 19.4% and each provides 30.5 kWe of net electrical power to the load. A He-Xe gas mixture serves as the turbine working fluid and cools the reactor core, entering at 904 K and exiting at 1149 K. Each CBC loop is coupled to a reactor sector, which is neutronically and thermally coupled, but hydraulically decoupled to the other two sectors, and to a NaK-78 secondary loop with two water heat pipes radiator panels. The segmented panels each consist of a forward fixed segment and two rear deployable segments, operating hydraulically in parallel. The deployed radiator has an effective surface area of 203 m2, and when the rear segments are folded, the stowed power system fits in the launch bay of the DELTA-IV Heavy launch vehicle. For enhanced reliability, the water heat pipes operate below 50% of their wicking limit; the sonic limit is not a concern because of the water, high vapor pressure at the temperatures of interest (384 - 491 K). The rejected power by the radiator peaks when the ratio of the lengths of evaporator sections of the longest and shortest heat pipes is the same as that of the major and minor widths of the segments. The shortest and hottest heat pipes in the rear segments operate at 491 K and 2.24 MPa, and each rejects 154 W. The longest heat pipes operate cooler (427 K and 0.52 MPa) and because they are 69% longer, reject more power (200 W each). The longest and hottest heat pipes in the forward segments reject the largest power (320 W each) while operating at ~ 46% of capillary limit. The vapor temperature and pressure in these heat pipes are 485 K and 1.97 MPa. By contrast, the

  10. High Temperature Water Heat Pipes Radiator for a Brayton Space Reactor Power System

    SciTech Connect

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2006-01-20

    A high temperature water heat pipes radiator design is developed for a space power system with a sectored gas-cooled reactor and three Closed Brayton Cycle (CBC) engines, for avoidance of single point failures in reactor cooling and energy conversion and rejection. The CBC engines operate at turbine inlet and exit temperatures of 1144 K and 952 K. They have a net efficiency of 19.4% and each provides 30.5 kWe of net electrical power to the load. A He-Xe gas mixture serves as the turbine working fluid and cools the reactor core, entering at 904 K and exiting at 1149 K. Each CBC loop is coupled to a reactor sector, which is neutronically and thermally coupled, but hydraulically decoupled to the other two sectors, and to a NaK-78 secondary loop with two water heat pipes radiator panels. The segmented panels each consist of a forward fixed segment and two rear deployable segments, operating hydraulically in parallel. The deployed radiator has an effective surface area of 203 m2, and when the rear segments are folded, the stowed power system fits in the launch bay of the DELTA-IV Heavy launch vehicle. For enhanced reliability, the water heat pipes operate below 50% of their wicking limit; the sonic limit is not a concern because of the water, high vapor pressure at the temperatures of interest (384 - 491 K). The rejected power by the radiator peaks when the ratio of the lengths of evaporator sections of the longest and shortest heat pipes is the same as that of the major and minor widths of the segments. The shortest and hottest heat pipes in the rear segments operate at 491 K and 2.24 MPa, and each rejects 154 W. The longest heat pipes operate cooler (427 K and 0.52 MPa) and because they are 69% longer, reject more power (200 W each). The longest and hottest heat pipes in the forward segments reject the largest power (320 W each) while operating at {approx} 46% of capillary limit. The vapor temperature and pressure in these heat pipes are 485 K and 1.97 MPa. By

  11. Review of the Tri-Agency Space Nuclear Reactor Power System Technology Program

    NASA Technical Reports Server (NTRS)

    Ambrus, J. H.; Wright, W. E.; Bunch, D. F.

    1984-01-01

    The Space Nuclear Reactor Power System Technology Program designated SP-100 was created in 1983 by NASA, the U.S. Department of Defense, and the Defense Advanced Research Projects Agency. Attention is presently given to the development history of SP-100 over the course of its first year, in which it has been engaged in program objectives' definition, the analysis of civil and military missions, nuclear power system functional requirements' definition, concept definition studies, the selection of primary concepts for technology feasibility validation, and the acquisition of initial experimental and analytical results.

  12. Flux stability and power control in the Soviet RBMK-1000 reactors

    SciTech Connect

    Meriwether, G.H.; McNeece, J.P.

    1993-08-01

    As a result of the Chernobyl accident, the Soviets have studied and implemented various design changes to improve the safety of the RBMK reactors. The safety measurements include modifications of the control rod configuration, fuel enrichment increase from 2.0 to 2.4 weight percent U-235, and installation of additional supplemental absorbers. The purpose of this study is to investigate the effects of increased fuel enrichment, different control rod positions, and supplemental absorber loadings on reactivity control, power distribution within the large RBMK core, and relative stability against power oscillations.

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

    SciTech Connect

    Samim Anghaie

    2002-08-13

    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

  14. Analysis and numerical optimization of gas turbine space power systems with nuclear fission reactor heat sources

    NASA Astrophysics Data System (ADS)

    Juhasz, Albert J.

    2005-07-01

    A new three objective optimization technique is developed and applied to find the operating conditions for fission reactor heated Closed Cycle Gas Turbine (CCGT) space power systems at which maximum efficiency, minimum radiator area, and minimum total system mass is achieved. Such CCGT space power systems incorporate a nuclear reactor heat source with its radiation shield; the rotating turbo-alternator, consisting of the compressor, turbine and the electric generator (three phase AC alternator); and the heat rejection subsystem, principally the space radiator, which enables the hot gas working fluid, emanating from either the turbine or a regenerative heat exchanger, to be cooled to compressor inlet conditions. Numerical mass models for all major subsystems and components developed during the course of this work are included in this report. The power systems modeled are applicable to future interplanetary missions within the Solar System and planetary surface power plants at mission destinations, such as our Moon, Mars, the Galilean moons (Io, Europa, Ganymede, and Callisto), or Saturn's moon Titan. The detailed governing equations for the thermodynamic processes of the Brayton cycle have been derived and successfully programmed along with the heat transfer processes associated with cycle heat exchangers and the space radiator. System performance and mass results have been validated against a commercially available non-linear optimization code and also against data from existing ground based power plants.

  15. Blue Ribbon Commission, Yucca Mountain Closure, Court Actions - Future of Decommissioned Reactors, Operating Reactors and Nuclear Power - 13249

    SciTech Connect

    Devgun, Jas S.

    2013-07-01

    Issues related to back-end of the nuclear fuel cycle continue to be difficult for the commercial nuclear power industry and for the decision makers at the national and international level. In the US, the 1982 NWPA required DOE to develop geological repositories for SNF and HLW but in spite of extensive site characterization efforts and over ten billion dollars spent, a repository opening is nowhere in sight. There has been constant litigation against the DOE by the nuclear utilities for breach of the 'standard contract' they signed with the DOE under the NWPA. The SNF inventory continues to rise both in the US and globally and the nuclear industry has turned to dry storage facilities at reactor locations. In US, the Blue Ribbon Commission on America's Nuclear Future issued its report in January 2012 and among other items, it recommends a new, consent-based approach to siting of facilities, prompt efforts to develop one or more geologic disposal facilities, and prompt efforts to develop one or more consolidated storage facilities. In addition, the March 2011 Fukushima Daiichi accident had a severe impact on the future growth of nuclear power. The nuclear industry is focusing on mitigation strategies for beyond design basis events and in the US, the industry is in the process of implementing the recommendations from NRC's Near Term Task Force. (authors)

  16. Quasi-heterogeneous efficient 3-D discrete ordinates CANDU calculations using Attila

    SciTech Connect

    Preeti, T.; Rulko, R.

    2012-07-01

    In this paper, 3-D quasi-heterogeneous large scale parallel Attila calculations of a generic CANDU test problem consisting of 42 complete fuel channels and a perpendicular to fuel reactivity device are presented. The solution method is that of discrete ordinates SN and the computational model is quasi-heterogeneous, i.e. fuel bundle is partially homogenized into five homogeneous rings consistently with the DRAGON code model used by the industry for the incremental cross-section generation. In calculations, the HELIOS-generated 45 macroscopic cross-sections library was used. This approach to CANDU calculations has the following advantages: 1) it allows detailed bundle (and eventually channel) power calculations for each fuel ring in a bundle, 2) it allows the exact reactivity device representation for its precise reactivity worth calculation, and 3) it eliminates the need for incremental cross-sections. Our results are compared to the reference Monte Carlo MCNP solution. In addition, the Attila SN method performance in CANDU calculations characterized by significant up scattering is discussed. (authors)

  17. Thermal and neutron-physical features of the nuclear reactor for a power pulsation plant for space applications

    NASA Astrophysics Data System (ADS)

    Gordeev, É. G.; Kaminskii, A. S.; Konyukhov, G. V.; Pavshuk, V. A.; Turbina, T. A.

    2012-05-01

    We have explored the possibility of creating small-size reactors with a high power output with the provision of thermal stability and nuclear safety under standard operating conditions and in emergency situations. The neutron-physical features of such a reactor have been considered and variants of its designs preserving the main principles and approaches of nuclear rocket engine technology are presented.

  18. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  19. 76 FR 8383 - Office of New Reactors; Interim Staff Guidance on Impacts of Construction of New Nuclear Power...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-02-14

    ..., ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (SRP),'' Chapter 1... COMMISSION Office of New Reactors; Interim Staff Guidance on Impacts of Construction of New Nuclear Power... Staff Guidance (ISG) COL-ISG-022 entitled ``Impacts of Construction of ] New Nuclear Power Plants...

  20. Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

    SciTech Connect

    Bernard, J.A. . Nuclear Reactor Lab.)

    1989-09-01

    This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power on spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.

  1. 100-kWe Lunar/Mars surface power utilizing the SP-100 reactor with dynamic conversion

    SciTech Connect

    Harty, R.B. ); Mason, L.S. )

    1993-01-15

    An integration study was performed coupling an SP-100 reactor with either a Brayton or Stirling power conversion subsystem. A power level of 100 kWe was selected for the study. The power system was to be compatible with both the lunar and Mars surface environment and require no site preparation. In addition, the reactor was to have integral shielding and be completely self-contained, including its own auxiliary power for start-up. Initial reliability studies were performed to determine power conversion redundancy and engine module size. Previous studies were used to select the power conversion optimum operating conditions (ratio of hot-side temperature to cold-side temperature). Results of the study indicated that either the Brayton or Stirling power conversion subsystems could be integrated with the SP-100 reactor for either a lunar or Mars surface power application. For the lunar environment, the reactor and primary coolant loop would be contained in a guard vessel to protect from a loss of primary loop containment. For the Mars environment, all refractory components including the reactor, primary coolant, and power conversion components would be contained in a vacuum vessel for protection against the CO[sub 2] environment. The vacuum would be maintained by an active ion pumping system. These active ion vacuum systems have no moving parts and have a long history of reliable operation.

  2. 100-kWe Lunar/Mars surface power utilizing the SP-100 reactor with dynamic conversion

    NASA Astrophysics Data System (ADS)

    Harty, Richard B.; Mason, Lee S.

    1993-01-01

    An integration study was performed coupling an SP-100 reactor with either a Brayton or Stirling power conversion subsystem. A power level of 100 kWe was selected for the study. The power system was to be compatible with both the lunar and Mars surface environment and require no site preparation. In addition, the reactor was to have integral shielding and be completely self-contained, including its own auxiliary power for start-up. Initial reliability studies were performed to determine power conversion redundancy and engine module size. Previous studies were used to select the power conversion optimum operating conditions (ratio of hot-side temperature to cold-side temperature). Results of the study indicated that either the Brayton or Stirling power conversion subsystems could be integrated with the SP-100 reactor for either a lunar or Mars surface power application. For the lunar environment, the reactor and primary coolant loop would be contained in a guard vessel to protect from a loss of primary loop containment. For the Mars environment, all refractory components including the reactor, primary coolant, and power conversion components would be contained in a vacuum vessel for protection against the CO2 environment. The vacuum would be maintained by an active ion pumping system. These active ion vacuum systems have no moving parts and have a long history of reliable operation.

  3. Development of fast breeder reactor fuel reprocessing technology at the Power Reactor and Nuclear Fuel Development Corporation

    SciTech Connect

    Kawata, T.; Takeda, H.; Togashi, A.; Hayashi, S. . Tokai Works); Stradley, J.G. )

    1991-01-01

    For the past two decades, a broad range of research development (R D) programs to establish fast breeder reactor (FBR) system and its associated fuel cycle technology have been pursued by the Power Reactor and Nuclear Fuel Development Corporation (PNC). Developmental activities for FBR fuel reprocessing technology have been primarily conducted at PNC Tokai Works where many important R D facilities for nuclear fuel cycle are located. These include cold and uranium tests for process equipment development in the Engineering Demonstration Facilities (EDF)-I and II, and laboratory-scale hot tests in the Chemical Processing Facility (CPF) where fuel dissolution and solvent extraction characteristics are being investigated with irradiated FBR fuel pins whose burn-up ranges up to 100,000 MWd/t. An extensive effort has also been made at EDF-III to develop advanced remote technology which enables to increase plant availability and to decrease radiation exposures to the workers in future reprocessing plants. The PNC and the United States Department of Energy (USDOE) entered into the joint collaboration in which the US shares the R Ds to support FBR fuel reprocessing program at the PNC. Several important R Ds on advanced process equipment such as a rotary dissolver and a centrifugal contactor system are in progress in a joint effort with the Oak Ridge National Laboratory (ORNL) Consolidated Fuel Reprocessing Program (CFRP). In order to facilitate hot testing on advanced processes and equipment, the design of a new engineering-scale hot test facility is now in progress aiming at the start of hot operation in late 90's. 31 refs., 2 tabs.

  4. Knowledges and abilities catalog for nuclear power plant operators: Savannah River Site (SRS) production reactors

    SciTech Connect

    Not Available

    1990-06-20

    The Knowledges and Abilities Catalog for Nuclear Power Plant Operations: Savannah River Site (SRS) Production Reactors, provides the basis for the development of content-valid certification examinations for Senior Reactor Operators (SROs) and Central Control Room Supervisors (SUP). The position of Shift Technical Engineer (STE) has been included in the catalog for completeness. This new SRS reactor operating shift crew position is held by an individual holding a CCR Supervisor Certification who has received special engineering and technical training. Also, the STE has a Bachelor of Science degree in engineering or a related technical field. The SRS catalog contains approximately 2500 knowledge and ability (K/A) statements for SROs and SUPs at heavy water moderated production reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring the health and safety of the public. The SRS K/A catalog is presently organized into five major sections: Plant Systems grouped by Safety Function, Plant Wide Generic K/As, Emergency Plant Evolutions, Theory and Components (to be developed).

  5. Environmental and safety assessment of LIBRA-SP: A light ion fusion power reactor design

    SciTech Connect

    Khater, H.Y.; Wittenberg, L.J.

    1996-12-31

    LIBRA-SP is a 1000 MWe light ion beam power reactor design study. The reactor structure is made of a low activation ferritic steel and uses LiPb as a breeder. The total activities in the blanket and reflector at shutdown are 721 MCi and 924 MCi, respectively. Hands-on maintenance is impossible anywhere inside the reactor chamber. The biological dose rates near the diode are too high at all times following shutdown allowing only for remote maintenance. The blanket and reflector could qualify for disposal as Class C low level waste. The dose to the maximally exposed individual in the vicinity of the reactor site due to the routine release of tritium is about 2.39 mrem/yr. Ten hours after a loss of coolant accident, the reflector produces a whole body (WB) early dose at the site boundary of 253 mrem. The blanket would produce a WB early dose of 8.91 rem. The potential off-site dose produced by the mobilization of LiPb during an accident is 142 mrem. A 100% release of the vulnerable tritium inventory present in the containment at any moment results in a WB early dose of 459 mrem. Release of the vulnerable tritium inventories present in the target factory and fuel reprocessing facility during an accident would result in WB early doses of 1.3 and 0.95 rem, respectively. 8 refs., 1 fig., 4 tabs.

  6. 10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Release of part of a power reactor facility or site for unrestricted use. 50.83 Section 50.83 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Transfers of Licenses-Creditors' Rights-Surrender of Licenses § 50.83 Release of part of a power reactor facility...

  7. An RF-powered micro-reactor for the detection of astrobiological target molecules on planetary bodies.

    PubMed

    Scott, Valerie J; Tse, Margaret; Shearn, Michael J; Siegel, Peter H; Amashukeli, Xenia

    2012-08-01

    We describe a sample-processing micro-reactor that utilizes 60 GHz RF radiation with approximately 730 mW of output power. The instrument design and performance characterization are described and then illustrated with modeling and experimental studies. The micro-reactor's efficiency on affecting hydrolysis of chemical bonds similar to those within large complex molecules was demonstrated: a disaccharide-sucrose-was hydrolyzed completely under micro-reactor conditions. The products of the micro-reactor-facilitated hydrolysis were analyzed using mass spectroscopy and proton nuclear magnetic resonance analytical techniques. PMID:22938313

  8. An alternative strategy for low specific power reactors to power interplanetary spacecraft, based on exploiting lasers and lunar resources

    SciTech Connect

    Logan, B.G.

    1989-02-02

    A key requirement setting the minimum electric propulsion performance (specific power ..cap alpha../sub e/ = kW/sub e//kg) for manned missions to Mars is the maximum allowable radiation dose to the crew during the long transits between Earth and Mars. Penetrating galactic cosmic rays and secondary neutron showers give about 0.1-rem/day dose, which only massive shielding (e.g., a meter of concrete) can reduce significantly. With a humane allowance for cabin space, the shielding mass becomes so large that it prohibitively escalates the propellant consumption required for reasonable trip times. This paper covers various proposed methods for using reactor power to propel spacecraft. 7 refs., 6 figs., 1 tab.

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

    NASA Technical Reports Server (NTRS)

    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.

  10. A Supercritical CO{sub 2} Cycle- a Promising Power Conversion System for Generation IV Reactors

    SciTech Connect

    Hejzlar, Pavel; Dostal, Vaclav; Driscoll, Michael J.

    2006-07-01

    Advances in power conversion systems (PCS) for Generation IV power plants are of high importance because of their impact on plant specific capital cost reduction, which can be more significant than the cost savings achieved through the modifications of the nuclear island itself. One such PCS candidate, especially attractive for reactor outlet temperatures in the range of 550 to 650 deg C, is applicable to lead-alloy, sodium, or liquid salt-cooled reactors, as well as direct-cycle CO{sub 2} cooled reactors. The efficiencies achievable in this medium temperature range exceed those of conventional Brayton cycles and supercritical steam Rankine cycles and are comparable to those of conventional helium Brayton cycles at turbine inlet temperatures of 800 to 900 deg C. The S-CO{sub 2} recompression cycle under evaluation at MIT, is described with its advantages, drawbacks and R and D needs. The cycle is shown to excel in efficiency, simplicity and compactness which projects to cost savings, and in lower sensitivity of efficiency to core bypass flow, component pressure losses and flow maldistribution in recuperators. On the other hand, the cycle is highly recuperative and thus requires very compact heat exchangers, poses challenges to design of piping for large units, and its control and part load operation is more complicated. (authors)

  11. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

    SciTech Connect

    REID, ROBERT S.; PEARSON, J. BOSIE; STEWART, ERIC T.

    2007-01-16

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  12. Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor

    SciTech Connect

    Pearson, J. Boise; Stewart, Eric T.; Reid, Robert S.

    2007-01-30

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  13. Ajustement du rechargement et des mecanismes de reactivite des reacteurs CANDU pour les cycles de combustible avances

    NASA Astrophysics Data System (ADS)

    St-Aubin, Emmanuel

    This research project main objectives are to set up and apply a methodology that can determine the potential of advanced thorium-based fuel cycles in CANDU reactors and that is able to adjust reactivity devices, in such a way as to maintain their reference efficiency for these new fuels. In order to select these fuel cycles, a large alternative fuel envelope is submitted to several discriminating criteria. A coarse parametric core modeling, that takes into account standard reactivity devices, is first used to highlight candidates presenting the best economical performances and to eliminate non viable options. Then, for the best candidates, the neutronic modeling is optimized before considering reactivity devices adjustment. For every reactivity device managed by the reactor regulating system, innovative generic optimization methods are used to achieve specific objectives for every fuel cycle, all of them being based on the reference natural uranium cycle behavior. Specific optimization objectives are assessed by simulating advanced fuel cycle for specific operating conditions, including : normal on-power refueling period, spurious reactor trip and fueling machine unavailibility. Unlike the generalized perturbative approach proposed in the OPTEX code, we have successfully implemented a multi-step method able to maximize both the energy extracted from the fuel using an equilibrium refueling optimization, and the reactivity devices adequacy. We also propose new reactivity device supercell models that provides accurate reactor databases for a fraction of the computing cost usually needed using a full model with a similar spatial discretization. Our approach is verified by comparing our simulation results with results published in the literature for the reference fuel cycle. The methodology developed identified advanced fuel cycles, containing up to 60%v. thorium, thereby increasing resources utilization by more than 50% and multiplying the fuel average exit burn-up by

  14. Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

    SciTech Connect

    Marcille, T. F.; Poston, D. I.; Kapernick, R. J.; Dixon, D. D.; Fischer, G. A.; Doherty, S. P.

    2006-01-20

    In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

  15. Investigation of applications for high-power, self-critical fissioning uranium plasma reactors

    NASA Technical Reports Server (NTRS)

    Rodgers, R. J.; Latham, T. S.; Krascella, N. L.

    1976-01-01

    Analytical studies were conducted to investigate potentially attractive applications for gaseous nuclear cavity reactors fueled by uranium hexafluoride and its decomposition products at temperatures of 2000 to 6000 K and total pressures of a few hundred atmospheres. Approximate operating conditions and performance levels for a class of nuclear reactors in which fission energy removal is accomplished principally by radiant heat transfer from the high temperature gaseous nuclear fuel to surrounding absorbing media were determined. The results show the radiant energy deposited in the absorbing media may be efficiently utilized in energy conversion system applications which include (1) a primary energy source for high thrust, high specific impulse space propulsion, (2) an energy source for highly efficient generation of electricity, and (3) a source of high intensity photon flux for heating working fluid gases for hydrogen production or MHD power extraction.

  16. Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

    NASA Astrophysics Data System (ADS)

    Marcille, T. F.; Dixon, D. D.; Fischer, G. A.; Doherty, S. P.; Poston, D. I.; Kapernick, R. J.

    2006-01-01

    In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

  17. Space reactor/Stirling cycle systems for high power lunar applications

    NASA Technical Reports Server (NTRS)

    Schmitz, Paul C.; Mason, Lee S.

    1991-01-01

    An analysis is performed to mathematically model a 550 kWe lunar base power supply which uses a SP-100 reactor coupled with Stirling converters. The reactor is placed in an excavation to keep activated coolant in the hole and to allow maintance of the components outside the hole. Two technology levels are considered. They are 1050 and 1300 K heater head Stirling converts. It is found that for a 1050 K converter the total mass which provided 1000 volts dc at 250 m is 14,366 kg while the 1300 K system mass is 12,104 kg. The radiation area of the 1050 and 1300 K systems are 641 and 356 sq m respectively. Comparisons are made with Brayton and thermionic systems with both near term and advanced technology considered.

  18. Space reactor/Stirling cycle systems for high power lunar application

    NASA Technical Reports Server (NTRS)

    Schmitz, Paul C.; Mason, Lee S.

    1991-01-01

    An analysis is performed to mathematically model a 550 kWe lunar base power supply which uses a SP-100 reactor coupled with Stirling converters. The reactor is placed in an excavation to keep activated coolant in the hole and to allow maintenance of the components outside the hole. Two technology levels are considered. They are 1050 and 1300 K heater head Stirling converts. It is found that for a 1050 K converter the total mass which provided 1000 volts DC at 250 m is 14,366 kg while the 1300 K system mass is 12,104 kg. The radiation area of the 1050 and 1300 K systems are 641 and 356 sq m respectively. Comparisons are made with Brayton and thermionic systems with both near term and advanced technology considered.

  19. Feasibility study of the University of Utah TRIGA reactor power upgrade in respect to control rod system

    NASA Astrophysics Data System (ADS)

    Cutic, Avdo

    The objectives of this thesis are twofold: to determine the highest achievable power levels of the current University of Utah TRIG Reactor (UUTR) core configuration with the existing three control rods, and to design the core for higher reactor power by optimizing the control rod worth. For the current core configuration, the maximum reactor power, eigenvalue keff, shutdown margin, and excess reactivity have been measured and calculated. These calculated estimates resulted from thermal power calibrations, and the control rod worth measurements at various power levels. The results were then used as a benchmark to verify the MCNP5 core simulations for the current core and then to design a core for higher reactor power. This study showed that the maximum achievable power with the current core configuration and control rod system is 150kW, which is 50kW higher than the licensed power of the UUTR. The maximum achievable UUTR core power with the existing fuel is determined by optimizing the core configuration and control rod worth, showing that a power upgrade of 500 kW is achievable. However, it requires a new control rod system consisting of a total of four control rods. The cost of such an upgrade is $115,000.

  20. Post 9-11 Security Issues for Non-Power Reactor Facilities

    SciTech Connect

    Zaffuts, P. J.

    2003-02-25

    This paper addresses the legal and practical issues arising out of the design and implementation of a security-enhancement program for non power reactor nuclear facilities. The security enhancements discussed are derived from the commercial nuclear power industry's approach to security. The nuclear power industry's long and successful experience with protecting highly sensitive assets provides a wealth of information and lessons that should be examined by other industries contemplating security improvements, including, but not limited to facilities using or disposing of nuclear materials. This paper describes the nuclear industry's approach to security, the advantages and disadvantages of its constituent elements, and the legal issues that facilities will need to address when adopting some or all of these elements in the absence of statutory or regulatory requirements to do so.

  1. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

  2. A 48-month extended fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

    SciTech Connect

    Erighin, M. A.

    2012-07-01

    The B and W mPower{sup TM} reactor is a small, rail-shippable pressurized water reactor (PWR) with an integral once-through steam generator and an electric power output of 150 MW, which is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height, but otherwise standard, PWR assemblies with the familiar 17 x 17 fuel rod array on a 21.5 cm inter-assembly pitch. The B and W mPower core design and cycle management plan, which were performed using the Studsvik core design code suite, follow the pattern of a typical nuclear reactor fuel cycle design and analysis performed by most nuclear fuel management organizations, such as fuel vendors and utilities. However, B and W is offering a core loading and cycle management plan for four years of continuous power operations without refueling and without the hurdles of chemical shim. (authors)

  3. Supercritical Carbon Dioxide Brayton Power Conversion Cycle Design for Optimized Battery-Type Integral Reactor System

    SciTech Connect

    Kim, Won J.; Kim, Tae W.; Sohn, Myoung S.; Suh, Kune Y.

    2006-07-01

    Supercritical carbon dioxide (SCO{sub 2}) promises a high power conversion efficiency of the recompression Brayton cycle due to its excellent compressibility reducing the compression work at the bottom of the cycle and to a higher density than helium or steam decreasing the component size. Therefore, the high SCO{sub 2} Brayton cycle efficiency as high as 45 % furnishes small sized nuclear reactors with economical benefits on the plant construction and maintenance. A 23 MWth BORIS (Battery Optimized Reactor Integral System) is being developed as a multipurpose reactor. BORIS, an integral-type optimized fast reactor with an ultra long life core, is coupled to the SCO{sub 2} Brayton cycle needing less room relative to the Rankine steam cycle because of its smaller components. The SCO{sub 2} Brayton cycle of BORIS consists of a 16 MW turbine, a 32 MW high temperature recuperator, a 14 MW low temperature recuperator, an 11 MW pre-cooler and 2 and 2.8 MW compressors. Entering six heat exchangers between primary and secondary system at 19.9 MPa and 663 K, the SCO{sub 2} leaves the heat exchangers at 19.9 MPa and 823 K. The promising secondary system efficiency of 45 % was calculated by a theoretical method in which the main parameters include pressure, temperature, heater power, the turbine's, recuperators' and compressors' efficiencies, and the flow split ratio of SCO{sub 2} going out from the low temperature recuperator. Test loop SOLOS (Shell-and-tube Overall Layout Optimization Study) is utilized to develop advanced techniques needed to adopt the shell-and-tube type heat exchanger in the secondary loop of BORIS by studying the SCO{sub 2} behavior from both thermal and hydrodynamic points of view. Concurrently, a computational fluid dynamics (CFD) code analysis is being conducted to develop an optimal analytical method of the SCO{sub 2} turbine efficiency having the parameters of flow characteristics of SCO{sub 2} passing through buckets of the turbine. These

  4. Performance Analyses of 38 kWe Turbo-Machine Unit for Space Reactor Power Systems

    SciTech Connect

    Gallo, Bruno M.; El-Genk, Mohamed S.

    2008-01-21

    This paper developed a design and investigated the performance of 38 kWe turbo-machine unit for space nuclear reactor power systems with Closed Brayton Cycle (CBC) energy conversion. The compressor and turbine of this unit are scaled versions of the NASA's BRU developed in the sixties and seventies. The performance results of turbo-machine unit are calculated for rotational speed up to 45 krpm, variable reactor thermal power and system pressure, and fixed turbine and compressor inlet temperatures of 1144 K and 400 K. The analyses used a detailed turbo-machine model developed at University of New Mexico that accounts for the various energy losses in the compressor and turbine and the effect of compressibility of the He-Xe (40 mole/g) working fluid with increased flow rate. The model also accounts for the changes in the physical and transport properties of the working fluid with temperature and pressure. Results show that a unit efficiency of 24.5% is achievable at rotation speed of 45 krpm and system pressure of 0.75 MPa, assuming shaft and electrical generator efficiencies of 86.7% and 90%. The corresponding net electric power output of the unit is 38.5 kWe, the flow rate of the working fluid is 1.667 kg/s, the pressure ratio and polytropic efficiency for the compressor are 1.60 and 83.1%, and 1.51 and 88.3% for the turbine.

  5. Performance Analyses of 38 kWe Turbo-Machine Unit for Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    Gallo, Bruno M.; El-Genk, Mohamed S.

    2008-01-01

    This paper developed a design and investigated the performance of 38 kWe turbo-machine unit for space nuclear reactor power systems with Closed Brayton Cycle (CBC) energy conversion. The compressor and turbine of this unit are scaled versions of the NASA's BRU developed in the sixties and seventies. The performance results of turbo-machine unit are calculated for rotational speed up to 45 krpm, variable reactor thermal power and system pressure, and fixed turbine and compressor inlet temperatures of 1144 K and 400 K. The analyses used a detailed turbo-machine model developed at the University of New Mexico that accounts for the various energy losses in the compressor and turbine and the effect of compressibility of the He-Xe (40 mole/g) working fluid with increased flow rate. The model also accounts for the changes in the physical and transport properties of the working fluid with temperature and pressure. Results show that a unit efficiency of 24.5% is achievable at rotation speed of 45 krpm and system pressure of 0.75 MPa, assuming shaft and electrical generator efficiencies of 86.7% and 90%. The corresponding net electric power output of the unit is 38.5 kWe, the flow rate of the working fluid is 1.667 kg/s, the pressure ratio and polytropic efficiency for the compressor are 1.60 and 83.1%, and 1.51 and 88.3% for the turbine.

  6. Experiments and models of general corrosion and flow-assisted corrosion of materials in nuclear reactor environments

    NASA Astrophysics Data System (ADS)

    Cook, William Gordon

    Corrosion and material degradation issues are of concern to all industries. However, the nuclear power industry must conform to more stringent construction, fabrication and operational guidelines due to the perceived additional risk of operating with radioactive components. Thus corrosion and material integrity are of considerable concern for the operators of nuclear power plants and the bodies that govern their operations. In order to keep corrosion low and maintain adequate material integrity, knowledge of the processes that govern the material's breakdown and failure in a given environment are essential. The work presented here details the current understanding of the general corrosion of stainless steel and carbon steel in nuclear reactor primary heat transport systems (PHTS) and examines the mechanisms and possible mitigation techniques for flow-assisted corrosion (FAC) in CANDU outlet feeder pipes. Mechanistic models have been developed based on first principles and a 'solution-pores' mechanism of metal corrosion. The models predict corrosion rates and material transport in the PHTS of a pressurized water reactor (PWR) and the influence of electrochemistry on the corrosion and flow-assisted corrosion of carbon steel in the CANDU outlet feeders. In-situ probes, based on an electrical resistance technique, were developed to measure the real-time corrosion rate of reactor materials in high-temperature water. The probes were used to evaluate the effects of coolant pH and flow on FAC of carbon steel as well as demonstrate of the use of titanium dioxide as a coolant additive to mitigated FAC in CANDU outlet feeder pipes.

  7. Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant

    SciTech Connect

    E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

    2008-08-01

    A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540°C and 900°C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

  8. EOIL power scaling in a 1-5 kW supersonic discharge-flow reactor

    NASA Astrophysics Data System (ADS)

    Davis, Steven J.; Lee, Seonkyung; Oakes, David B.; Haney, Julie; Magill, John C.; Paulsen, Dwane A.; Cataldi, Paul; Galbally-Kinney, Kristin L.; Vu, Danthu; Polex, Jan; Kessler, William J.; Rawlins, Wilson T.

    2008-02-01

    Scaling of EOIL systems to higher powers requires extension of electric discharge powers into the kW range and beyond with high efficiency and singlet oxygen yield. We have previously demonstrated a high-power microwave discharge approach capable of generating singlet oxygen yields of ~25% at ~50 torr pressure and 1 kW power. This paper describes the implementation of this method in a supersonic flow reactor designed for systematic investigations of the scaling of gain and lasing with power and flow conditions. The 2450 MHz microwave discharge, 1 to 5 kW, is confined near the flow axis by a swirl flow. The discharge effluent, containing active species including O II(a1Δ g, b1Σ g +), O( 3P), and O 3, passes through a 2-D flow duct equipped with a supersonic nozzle and cavity. I2 is injected upstream of the supersonic nozzle. The apparatus is water-cooled, and is modular to permit a variety of inlet, nozzle, and optical configurations. A comprehensive suite of optical emission and absorption diagnostics is used to monitor the absolute concentrations of O II(a), O II(b), O( 3P), O 3, I II, I(2P 3/2), I(2P 1/2), small-signal gain, and temperature in both the subsonic and supersonic flow streams. We discuss initial measurements of singlet oxygen and I* excitation kinetics at 1 kW power.

  9. The scalability of OTR (out-of-core thermionic reactor) space nuclear power systems

    SciTech Connect

    Gallup, D.R.

    1990-03-01

    In this document, masses of the STAR-C power system and an optimized out-of-core thermionic reactor (OTR) power system versus power level are investigated. The impacts of key system parameters on system performance are also addressed. The STAR-C is mass competitive below about 15 kWe, but at higher power levels the scalability is relatively poor. An optimized OR is the least massive space nuclear power system below 25 kWe, and scales well to 50 kWe. The system parameters that have a significant impact on the scalability of the STAR-C are core thermal flux, thermionic converter efficiency, and core length to diameter ratio. The emissivity of the core surface is shown to be a relatively unimportant parameter. For an optimized OR power system, the most significant system parameter is the maximum allowable fuel temperature. It is also shown that if advanced radiation-hardened electronics are used in the satellite payload, a very large mass savings is realized. 10 refs., 23 figs., 7 tabs.

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

    NASA Technical Reports Server (NTRS)

    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.

  11. Compatibility tests of materials for a lithium-cooled space power reactor concept

    NASA Technical Reports Server (NTRS)

    Sinclair, J. H.

    1973-01-01

    Materials for a lithium-cooled space power reactor concept must be chemically compatible for up to 50,000 hr at high temperature. Capsule tests at 1040 C (1900 F) were made of material combinations of prime interest: T-111 in direct contact with uranium mononitride (UN), Un in vacuum separated from T-111 by tungsten wire, UN with various oxygen impurity levels enclosed in tungsten wire lithium-filled T-111 capsules, and TZM and lithium together in T-111 capsules. All combinations were compatible for over 2800 hr except for T-111 in direct contact with UN.

  12. Source-term reevaluation for US commercial nuclear power reactors: a status report

    SciTech Connect

    Herzenberg, C.L.; Ball, J.R.; Ramaswami, D.

    1984-12-01

    Only results that had been discussed publicly, had been published in the open literature, or were available in preliminary reports as of September 30, 1984, are included here. More than 20 organizations are participating in source-term programs, which have been undertaken to examine severe accident phenomena in light-water power reactors (including the chemical and physical behavior of fission products under accident conditions), update and reevaluate source terms, and resolve differences between predictions and observations of radiation releases and related phenomena. Results from these source-term activities have been documented in over 100 publications to date.

  13. Collaborative investigations of in-service irradiated material from the Japan Power Demonstration Reactor pressure vessel

    SciTech Connect

    Corwin, W.R.; Broadhead, B.L.; Suzuki, M.; Kohsaka, A.

    1997-02-01

    There is a need to validate the results of irradiation effects research by the examination of material taken directly from the wall of a pressure vessel that has been irradiated during normal service. Just such an evaluation is currently being conducted on material from the wall of the pressure vessel from the Japan Power Demonstration Reactor (JPDR). The research is being jointly performed at the Tokai Research Establishment of the Japan Atomic Energy Research Institute (JAERI) and by the Nuclear Regulatory Commission (NRC)-funded Heavy-Section Steel Irradiation Program at the Oak Ridge National Laboratory (ORNL).

  14. Assessments of Longevity of Equipment Metal of Nuclear Power Plants equipped with Reactors VVER-1000

    SciTech Connect

    Gorbatykh, V.P.; Al Kassem, S.N.

    2004-07-01

    Characteristics of damage processes of metal of coffer-dams of steam generators collectors at nuclear power plants (NPPs) equipped with reactors VVER-1000 have been mentioned; principles of construction of longevity function has been cited and new approach has been shown while solving the problem of the longevity of the metal resource by substantiating the technological actions with new mode characteristics, performed with the help of specially developed equations and formulae, where practically all damage processes and all influencing factors can be accounted. (authors)

  15. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    SciTech Connect

    1995-08-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  16. PR-EDB: Power Reactor Embrittlement Data Base, Version 2. Revision 2, Program description

    SciTech Connect

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.; Taylor, B.J.

    1994-01-01

    Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes Standard Review Plans (SRP`s) and Guides for license renewal can be greatly expedited by the use of a well-designed computerized data base. Also, such a data base is essential for the validation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current version of the PR-EDB contains the Charpy test data that were irradiated in 252 capsules of 96 reactors and consists of 207 data points for heat-affected-zone (HAZ) materials (98 different HAZ), 227 data points for weld materials (105 different welds), 524 data points for base materials (136 different base materials), including 297 plate data points (85 different plates), 119 forging data points (31) different forging), and 108 correlation monitor materials data points (3 different plates). The data files are given in dBASE format and can be accessed with any computer using the DOS operating system. ``User-friendly`` utility programs are used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in Appendix D.

  17. Plutonium disposition in the BN-600 fast-neutron reactor at the Beloyarsk nuclear power plant

    NASA Astrophysics Data System (ADS)

    Moses, D. L.; Chebeskov, A. N.; Matveev, V. I.; Vasiliev, B. A.; Maltsev, V. V.

    In 1996, the United States and the Russian Federation completed an initial joint study that evaluated the candidate options for the disposition of surplus weapons-derived plutonium in both countries. While Russia advocates building new reactors for converting weapons-derived plutonium to spent fuel, the cost is high, and the continuing joint study of the Russian options is considering only the use of the existing VVER-1000 LWRs in Russia (and possibly in Ukraine) and the existing BN-600 fast-neutron reactor at the Beloyarsk Nuclear Power Plant in Russia. The BN-600 reactor, which currently uses enriched uranium fuel, is capable with certain design modifications of converting up to 1.3 metric tons (MT) of surplus weapons-derived plutonium to spent fuel each year. The steps needed to convert BN-600 to a plutonium-burner core will be discussed. The step involving the hybrid core allows an early and timely start that takes advantage of the limited capacity for fabricating uranium-plutonium mixed-oxide fuel early in the disposition program. The design lifetime of BN-600 must safely and reliably be extended by 10 yr to at least 2020 so that a sufficient amount of plutonium (˜20 MT) can be converted to spent fuel.

  18. New concept of small power reactor without on-site refueling for non-proliferation

    SciTech Connect

    Brown, N.W., LLNL

    1998-07-13

    Energy demand in developing countries is increasing to support growing populations and economies. This demand is expected to continue growing at a rapid pace well into the next century. Because current power sources, including fossil, renewable, and nuclear, cannot meet energy demands, many developing countries are interested in building a new generation of small reactor systems to help meet their needs. The U.S. recognizes the need for energy in the developing countries. In its 1998 Comprehensive Energy Strategy, the Department of Energy calls for research into low-cost, proliferation- resistant, nuclear reactor technologies to ensure that this demand can be met in a manner consistent with U.S. non-proliferation goals and policies. This research has two primary thrusts: first, the development of a small proliferation-resistant nuclear system (i.e., a technology focus); second, the continuation of open communication with the international community through early engagement and cooperation on small reactor development. A system that meets developing country requirements must: (1) achieve reliably safe operation with a minimum of maintenance and supporting infrastructure; (2) offer economic competitiveness with alternative energy sources available to the candidate sites; and (3) demonstrate significant improvements in proliferation resistance relative to existing reactor systems. These challenges are the most significant driving forces behind the LLNL proposed program for development of a new, small nuclear reactor system. This report describes a technical approach for developing small nuclear power systems for use in developing countries. The approach being proposed will establish a preliminary set of requirements that, if met, will cause new innovative approaches to system design to be used. The proposed approach will borrow from experience gained over the past forty years with four types of nuclear reactor technologies (LWR, LMR, HTGR, and MSR) to develop four or

  19. Dynamic parameters test of Haiyang Nuclear Power Engineering in reactor areas, China

    NASA Astrophysics Data System (ADS)

    Zhou, N.; Zhao, S.; Sun, L.

    2012-12-01

    Haiyang Nuclear Power Project is located in Haiyang city, China. It consists of 6×1000MW AP1000 Nuclear Power generator sets. The dynamic parameters of the rockmass are essential for the design of the nuclear power plant. No.1 and No.2 reactor area are taken as research target in this paper. Sonic logging, single hole and cross-hole wave velocity are carried out respectively on the site. There are four types of rock lithology within the measured depth. They are siltstone, fine sandstone, shale and allgovite. The total depth of sonic logging is 409.8m and 2049 test points. The sound wave velocity of the rocks are respectively 5521 m/s, 5576m/s, 5318 m/s and 5576 m/s. Accroding to the statistic data, among medium weathered fine sandstone, fairly broken is majority, broken and relatively integrity are second, part of integrity. Medium weathered siltstone, relatively integrity is mojority, fairly broken is second. Medium weathered shale, fairly broken is majority, broken and relatively integrity for the next and part of integrity. Slight weathered fine sandstone, siltstone, shale and allgovite, integrity is the mojority, relatively integrity for the next, part of fairly broken.The single hole wave velocity tests are set in two boreholesin No.1 reactor area and No.2 reactor area respectively. The test depths of two holes are 2-24m, and the others are 2-40m. The wave velocity data are calculated at different depth in each holes and dynamic parameters. According to the test statistic data, the wave velocity and the dynamic parameter values of rockmass are distinctly influenced by the weathering degree. The test results are list in table 1. 3 groups of cross hole wave velocity tests are set for No.1 and 2 reactor area, No.1 reactor area: B16, B16-1, B20(Direction:175°, depth: 100m); B10, B10-1, B11(269°, 40m); B21, B21-1, B17(154°, 40m); with HB16, HB10, HB21 as trigger holes; No.2 reactor area: B47, B47-1, HB51(176°, 100m); B40, B40-1, B41(272°, 40m); B42, B42-1, B

  20. On the power and size of tokamak fusion pilot plants and reactors

    NASA Astrophysics Data System (ADS)

    Costley, A. E.; Hugill, J.; Buxton, P. F.

    2015-03-01

    It is generally accepted that the route to fusion power involves large devices of ITER scale or larger. However, we show, contrary to expectations, that for steady state tokamaks operating at fixed fractions of the density and beta limits, the fusion gain, Qfus, depends mainly on the absolute level of the fusion power and the energy confinement, and only weakly on the device size. Our investigations are carried out using a system code and also by analytical means. Further, we show that for the two qualitatively different global scalings that have been developed to fit the data contained in the ITER ELMy H-mode database, i.e. the normally used beta-dependent IPB98y2 scaling and the alternative beta-independent scalings, the power needed for high fusion performance differs substantially, typically by factors of three to four. Taken together, these two findings imply that lower power, smaller, and hence potentially lower cost, pilot plants and reactors than currently envisaged may be possible. The main parameters of a candidate low power (˜180 MW), high Qfus (˜5), relatively small (˜1.35 m major radius) device are given.

  1. SP-100, the US Space Nuclear Reactor Power Program. Technical information report

    SciTech Connect

    Truscello, V.C.

    1983-11-01

    DARPA, in conjunction with DOE`s Office of Nuclear Energy, and NASA`s Office of Aeronautics and Space Technology are jointly sponsoring a space nuclear reactor power system program known as the Space Power-100 (SP-100) Development Project. The program is presently in the critical technology phase. This phase, better known as technology assessment and advancement, includes mission requirements definition, system conceptual designs, and critical technology development. A ground test phase decision is scheduled for July 1985. If the decision is positive, the next phase would begin in fiscal year 1986. An overriding concern in conducting this program is to ensure that nuclear safety is being properly addressed even in these early stages.

  2. Experimental Evaluation of a Water Shield for a Surface Power Reactor

    NASA Technical Reports Server (NTRS)

    Pearson, J. B.; Reid, R.; Sadasivan, P.; Stewart, E.

    2007-01-01

    A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. A representative lunar surface reactor design is evaluated at various power levels in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The evaluation compares the experimental data from the WST to CFD models. Performance of a water shield on the lunar surface is predicted by CFD models anchored to test data, and by matching relevant dimensionless parameters.

  3. An assessment and validation study of nuclear reactors for low power space applications

    NASA Technical Reports Server (NTRS)

    Klein, A. C.; Gedeon, S. R.; Morey, D. C.

    1987-01-01

    The feasibility and safety of six conceptual small, low power nuclear reactor designs was evaluated. Feasibility evaluations included the determination of sufficient reactivity margins for seven years of full power operation and safe shutdown as well as handling during pre-launch assembly phases. Safety evaluations were concerned with the potential for maintaining subcritical conditions in the event of launch or transportation accidents. These included water immersion accident scenarios both with and without water flooding the core. Results show that most of the concepts can potentially meet the feasibility and safety requirements; however, due to the preliminary nature of the designs considered, more detailed designs will be necessary to enable these concepts to fully meet the safety requirements.

  4. Test Results From a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    NASA Technical Reports Server (NTRS)

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

    2009-01-01

    The Brayton Power Conversion Unit (BPCU) located at NASA Glenn Research Center (GRC) in Cleveland, OH is a closed cycle system incorporating a turboaltemator, recuperator, and gas cooler connected by gas ducts to an external gas heater. For this series of tests, the BPCU was modified by replacing the gas heater with the Direct Drive Gas heater or DOG. The DOG uses electric resistance heaters to simulate a fast spectrum nuclear reactor similar to those proposed for space power applications. The combined system thermal transient behavior was the focus of these tests. The BPCU was operated at various steady state points. At each point it was subjected to transient changes involving shaft rotational speed or DOG electrical input. This paper outlines the changes made to the test unit and describes the testing that took place along with the test results.

  5. Office of Analysis and Evaluation of Operational Data 1989 annual report, Power reactors

    SciTech Connect

    1990-07-01

    The annual report of the US Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1989. The report is published in two separate parts. This document, NUREG-1272, Vol. 4, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about the trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year and summarizes information from such sources as licensee event reports, diagnostic evaluations, and reports to the NRC's Operations Center. This report also compiles the status of staff actions resulting from previous Incident Investigation Team (IIT) reports. 16 figs., 9 tabs.

  6. Medium-Power Lead-Alloy Fast Reactor Balance-of-Plant Options

    SciTech Connect

    Dostal, Vaclav; Hejzlar, Pavel; Todreas, Neil E.; Buongiorno, Jacopo

    2004-09-15

    Proper selection of the power conversion cycle is a very important step in the design of a nuclear reactor. Due to the higher core outlet temperature ({approx}550 deg. C) compared to that of light water reactors ({approx}300 deg. C), a wide portfolio of power cycles is available for the lead alloy fast reactor (LFR). Comparison of the following cycles for the LFR was performed: superheated steam (direct and indirect), supercritical steam, helium Brayton, and supercritical CO{sub 2} (S-CO{sub 2}) recompression. Heat transfer from primary to secondary coolant was first analyzed and then the steam generators or heat exchangers were designed. The direct generation of steam in the lead alloy coolant was also evaluated. The resulting temperatures of the secondary fluids are in the range of 530-545 deg. C, dictated by the fixed space available for the heat exchangers in the reactor vessel. For the direct steam generation situation, the temperature is 312 deg. C. Optimization of each power cycle was carried out, yielding net plant efficiency of around 40% for the superheated steam cycle while the supercritical steam and S-CO{sub 2} cycles achieved net plant efficiency of 41%. The cycles were then compared based on their net plant efficiency and potential for low capital cost. The superheated steam cycle is a very good candidate cycle given its reasonably high net plant efficiency and ease of implementation based on the extensive knowledge and operating experience with this cycle. Although the supercritical steam cycle net plant efficiency is slightly better than that of the superheated steam cycle, its high complexity and high pressure result in higher capital cost, negatively affecting plant economics. The helium Brayton cycle achieves low net plant efficiency due to the low lead alloy core outlet temperature, and therefore, even though it is a simpler cycle than the steam cycles, its performance is mediocre in this application. The prime candidate, however, appears to be

  7. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

    NASA Technical Reports Server (NTRS)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

    1996-01-01

    Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

  8. Technical evaluation of major candidate blanket systems for fusion power reactor

    NASA Astrophysics Data System (ADS)

    Tone, Tatsuzo; Seki, Masahiro; Minato, Akio

    1987-03-01

    The key functions required for tritium breeding blankets for a fusion power reactor are ; (1) self-sufficient tritium breeding, (2) in-situ tritium recovery and low tritium inventory, (3) high temperature cooling giving a high efficiency of electricity generation and (4) thermo-mechanical reliability and simplified remote maintenance to obtain high plant availability. Blanket performance is substantially governed by materials selection. Major options of structure/breeder/coolant/neutron multiplier materials considered for the present design study are PCA/Li/sub 2/O/H/sub 2/O/Be, Mo-alloy/Li/sub 2/O/He/Be, Mo-alloy/LiAlO/sub 2//He/Be, V-alloy/Li/Li/none, and Mo-alloy/Li/He/none. In addition, remote maintenance of blankets, tritium recovery system, heat transport and energy conversion have been investigated. In this report, technological problems and critical R and D issues for power reactor blanket development are identified and a comparison of major candidate blanket concepts is discussed in terms of the present materials data base, economic performance, prospects for future improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies. improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies.

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

    NASA Technical Reports Server (NTRS)

    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.

  10. Transmutation and activation analysis for divertor materials in a HCLL-type fusion power reactor

    NASA Astrophysics Data System (ADS)

    Fischer, U.; Pereslavtsev, P.; Möslang, A.; Rieth, M.

    2009-04-01

    The activation and transmutation of tungsten and tantalum as plasma facing materials was assessed for a helium cooled divertor irradiated in a typical fusion power reactor based on the use of Helium-cooled Lithium Lead (HCLL) blankets. 3D activation calculations were performed by applying a programme system linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. Special attention was given to the proper treatment of the resonance shielding of tungsten and tantalum by using reaction rates provided directly by MCNP on the basis of continuous energy activation cross-section data. It was shown that the long-term activation behaviour is dominated by activation products of the assumed tramp material while the short-term behaviour is due to the activation of the stable Ta and W isotopes. The recycling limit for remote handling of 100 mSv/h can be achieved after decay times of 10 and 50 years for Ta and W, respectively. The elemental transmutation rates of Ta and W were shown to be on a moderate level for the HCLL-type fusion power reactor.

  11. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  12. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

    SciTech Connect

    Mac Donald, Philip Elsworth

    2002-06-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  13. Time-Varying Characteristics Analysis and Fuzzy Controller Systematic Design Method for Pressurized Water Reactor Power Control

    SciTech Connect

    Liu Shengzhi; Zhang Naiyao; Cui Zhenhua

    2004-11-15

    In this paper a systematic design method of fuzzy control systems is applied to the pressurized water reactor's (PWR) power control. The paper includes three parts. In the first part, a simplified time-varying linear model of the PWR power system is constructed, and its inner structure and time-varying characteristics are analyzed. That provides a solid basis for study and design of the nuclear reactor power control system. In the second part, a systematic design method of fuzzy control systems is introduced and applied to control the nuclear reactor power process. The design procedures and parameters are given in detail. This systematic design method has some notable advantages. The control of a global fuzzy model can be decomposed into controlling a set of linear submodels. Each submodel controller can be independently designed by using a linear quadratic regulator approach. This systematic design method gives a sufficient and necessary condition to guarantee the stability of fuzzy control systems; thus, better control performance can be obtained due to the accurate control gains. In the third part, the control performance of the nuclear reactor fuzzy control system is examined by simulation experiments, including nuclear reactor power shutdown, start-up, and adjustment operations. The satisfactory experiment results have shown that the systematic design method for fuzzy control systems is effective and feasible.

  14. Combined on-board hydride slurry storage and reactor system and process for hydrogen-powered vehicles and devices

    SciTech Connect

    Brooks, Kriston P; Holladay, Jamelyn D; Simmons, Kevin L; Herling, Darrell R

    2014-11-18

    An on-board hydride storage system and process are described. The system includes a slurry storage system that includes a slurry reactor and a variable concentration slurry. In one preferred configuration, the storage system stores a slurry containing a hydride storage material in a carrier fluid at a first concentration of hydride solids. The slurry reactor receives the slurry containing a second concentration of the hydride storage material and releases hydrogen as a fuel to hydrogen-power devices and vehicles.

  15. Design and Development of a Robotic Crawler for CANDU Fuel Channel Inspection

    NASA Astrophysics Data System (ADS)

    Shukla, Shivam

    For the design of a new robotic crawler drive unit for CANDU fuel channel inspection, a complete design and screening process was done in order to fulfil the objective of this research. A brief explanation of CANDU reactors is provided along with a discussion of the inspection systems that are currently in use. A study of some existing inspection systems is presented which was used for the development of the new robotic crawler design. A number of concepts were generated which underwent a screening process with the help of two design tools. With the help of these tools, a concept was chosen as the final design and details of it are presented. To demonstrate a proof-of-concept, the physical prototype of the robotic crawler was manufactured and assembled. A speed controller was implemented in the final design of the robotic crawler. A set of test procedures were performed on the final design and the results are discussed. Some improvements that can be done on the final design of the robotic crawler are also discussed in the final section of this thesis.

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

    NASA Technical Reports Server (NTRS)

    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.

  17. Study on neutronic of very small Pb - Bi cooled no-onsite refueling nuclear power reactor (VSPINNOR)

    SciTech Connect

    Arianto, Fajar; Su'ud, Zaki; Zuhair

    2014-09-30

    A conceptual design study on Very Small Pb-Bi No-Onsite Refueling Cooled Nuclear Reactor (VSPINNOR) with Uranium nitride fuel using MCNPX program has been performed. In this design the reactor core is divided into three regions with different enrichment. At the center of the core is laid fuel without enrichment (internal blanket). While for the outer region using fuel enrichment variations. VSPINNOR fast reactor was operated for 10 years without refueling. Neutronic analysis shows optimized result of VSPINNOR has a core of 50 cm radius and 100 cm height with 300 MWth thermal power output at 60% fuel fraction that can be operated 18 years without refueling or fuel shuffling.

  18. Electron versus proton accelerator driven sub-critical system performance using TRIGA reactors at power

    SciTech Connect

    Carta, M.; Burgio, N.; D'Angelo, A.; Santagata, A.; Petrovich, C.; Schikorr, M.; Beller, D.; Felice, L. S.; Imel, G.; Salvatores, M.

    2006-07-01

    This paper provides a comparison of the performance of an electron accelerator-driven experiment, under discussion within the Reactor Accelerator Coupling Experiments (RACE) Project, being conducted within the U.S. Dept. of Energy's Advanced Fuel Cycle Initiative (AFCI), and of the proton-driven experiment TRADE (TRIGA Accelerator Driven Experiment) originally planned at ENEA-Casaccia in Italy. Both experiments foresee the coupling to sub-critical TRIGA core configurations, and are aimed to investigate the relevant kinetic and dynamic accelerator-driven systems (ADS) core behavior characteristics in the presence of thermal reactivity feedback effects. TRADE was based on the coupling of an upgraded proton cyclotron, producing neutrons via spallation reactions on a tantalum (Ta) target, with the core driven at a maximum power around 200 kW. RACE is based on the coupling of an Electron Linac accelerator, producing neutrons via photoneutron reactions on a tungsten-copper (W-Cu) or uranium (U) target, with the core driven at a maximum power around 50 kW. The paper is focused on analysis of expected dynamic power response of the RACE core following reactivity and/or source transients. TRADE and RACE target-core power coupling coefficients are compared and discussed. (authors)

  19. A Supercritical CO{sub 2} Gas Turbine Power Cycle for Next-Generation Nuclear Reactors

    SciTech Connect

    Dostal, Vaclav; Driscoll, Michael J.; Hejzlar, Pavel; Todreas, Neil E.

    2002-07-01

    Although proposed more than 35 years ago, the use of supercritical CO{sub 2} as the working fluid in a closed circuit Brayton cycle has so far not been implemented in practice. Industrial experience in several other relevant applications has improved prospects, and its good efficiency at modest temperatures (e.g., {approx}45% at 550 deg. C) make this cycle attractive for a variety of advanced nuclear reactor concepts. The version described here is for a gas-cooled, modular fast reactor. In the proposed gas-cooled fast breeder reactor design of present interest, CO{sub 2} is also especially attractive because it allows the use of metal fuel and core structures. The principal advantage of a supercritical CO{sub 2} Brayton cycle is its reduced compression work compared to an ideal gas such as helium: about 15% of gross power turbine output vs. 40% or so. This also permits the simplification of use of a single compressor stage without inter-cooling. The requisite high pressure ({approx}20 MPa) also has the benefit of more compact heat exchangers and turbines. Finally, CO{sub 2} requires significantly fewer turbine stages than He, its principal competitor for nuclear gas turbine service. One disadvantage of CO{sub 2} in a direct cycle application is the production of N-16, which will require turbine plant shielding (albeit much less than in a BWR). The cycle efficiency is also very sensitive to recuperator effectiveness and compressor inlet temperature. It was found necessary to split the recuperator into separate high-and low-temperature components, and to employ intermediate re-compression, to avoid having a pinch-point in the cold end of the recuperator. Over the past several decades developments have taken place that make the acceptance of supercritical CO{sub 2} systems more likely: supercritical CO{sub 2} pipelines are in use in the western US in oil-recovery operations; 14 advanced gas-cooled reactors (AGR) are employed in the UK at CO{sub 2} temperatures up to

  20. Introduction of water chemistry conditions of the secondary coolant circuit with metering organic amines at nuclear power stations equipped with VVER-1000 reactors

    NASA Astrophysics Data System (ADS)

    Tyapkov, V. F.; Erpyleva, S. F.; Bykova, V. V.

    2009-05-01

    Results from introduction of new water chemistry conditions involving metering of organic amines (morpholine and ethanolamine) at nuclear power stations equipped with VVER-1000 reactors are presented.

  1. POWER CYCLE AND STRESS ANALYSES FOR HIGH TEMPERATURE GAS-COOLED REACTOR

    SciTech Connect

    Oh, Chang H; Davis, Cliff; Hawkes, Brian D; Sherman, Steven R

    2007-05-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with three turbines and four compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with three stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and a 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles to

  2. Lunar Nuclear Power Plant With Solid Core Reactor, Heatpipes and Thermoelectric Conversion

    SciTech Connect

    Sayre, Edwin D.; Ring, Peter J.; Brown, Neil; Elsner, Norbert B.; Bass, John C.

    2008-01-21

    This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter and connect to the radiator. The solid core reactor is made from HT9 alloy. The fuel is uranium oxide with 90% enrichment. The thermoelectric converter is bonded to the outside of the 1.10 inch ID heat pipe and is 30 inches long. The thermoelectric couple is Si/SiGe-Si/SiC Quantum Well with over 20% efficiency with an 890 K hot side and a 490 K cold side and produces 625 Watts. 176 converters produce 110 kWe. With less than 10% loss in controls this yields 100 kWe for use. The cylindrical thermoelectric converter is designed and fabricated by HIPing to keep brittle materials in compression and to ensure conductivity. The solid core is fabricated by machining the heatpipe tubes with 6 grooves that are diffusion bonded together by HIPing to form the fuel tubes. The maximum temperature of the heat pipes is 940 K and the return flow temperature is 890 K. The reactor core is hexagonal shaped, 61 cm. wide and 76.2 cm high with 12 rotating control drums surrounding it. There is shielding to protect components and human habitation. The radiator is daisy shaped at 45 degrees with each petal 5.5 meters long. The design life is ten years.

  3. Innovative self-powered submersible microbial electrolysis cell (SMEC) for biohydrogen production from anaerobic reactors.

    PubMed

    Zhang, Yifeng; Angelidaki, Irini

    2012-05-15

    A self-powered submersible microbial electrolysis cell (SMEC), in which a specially designed anode chamber and external electricity supply were not needed, was developed for in situ biohydrogen production from anaerobic reactors. In batch experiments, the hydrogen production rate reached 17.8 mL/L/d at the initial acetate concentration of 410 mg/L (5 mM), while the cathodic hydrogen recovery ( [Formula: see text] ) and overall systemic coulombic efficiency (CE(os)) were 93% and 28%, respectively, and the systemic hydrogen yield ( [Formula: see text] ) peaked at 1.27 mol-H(2)/mol-acetate. The hydrogen production increased along with acetate and buffer concentration. The highest hydrogen production rate of 32.2 mL/L/d and [Formula: see text] of 1.43 mol-H(2)/mol-acetate were achieved at 1640 mg/L (20 mM) acetate and 100 mM phosphate buffer. Further evaluation of the reactor under single electricity-generating or hydrogen-producing mode indicated that further improvement of voltage output and reduction of electron losses were essential for efficient hydrogen generation. In addition, alternate exchanging the electricity-assisting and hydrogen-producing function between the two cell units of the SMEC was found to be an effective approach to inhibit methanogens. Furthermore, 16S rRNA genes analysis showed that this special operation strategy resulted same microbial community structures in the anodic biofilms of the two cell units. The simple, compact and in situ applicable SMEC offers new opportunities for reactor design for a microbial electricity-assisted biohydrogen production system. PMID:22402271

  4. A physical description of fission product behavior fuels for advanced power reactors.

    SciTech Connect

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  5. Lunar Nuclear Power Plant With Solid Core Reactor, Heatpipes and Thermoelectric Conversion

    NASA Astrophysics Data System (ADS)

    Sayre, Edwin D.; Ring, Peter J.; Brown, Neil; Elsner, Norbert B.; Bass, John C.

    2008-01-01

    This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter and connect to the radiator. The solid core reactor is made from HT9 alloy. The fuel is uranium oxide with 90% enrichment. The thermoelectric converter is bonded to the outside of the 1.10 inch ID heat pipe and is 30 inches long. The thermoelectric couple is Si/SiGe-Si/SiC Quantum Well with over 20% efficiency with an 890 K hot side and a 490 K cold side and produces 625 Watts. 176 converters produce 110 kWe. With less than 10% loss in controls this yields 100 kWe for use. The cylindrical thermoelectric converter is designed and fabricated by HIPing to keep brittle materials in compression and to ensure conductivity. The solid core is fabricated by machining the heatpipe tubes with 6 grooves that are diffusion bonded together by HIPing to form the fuel tubes. The maximum temperature of the heat pipes is 940 K and the return flow temperature is 890 K. The reactor core is hexagonal shaped, 61 cm. wide and 76.2 cm high with 12 rotating control drums surrounding it. There is shielding to protect components and human habitation. The radiator is daisy shaped at 45 degrees with each petal 5.5 meters long. The design life is ten years.

  6. NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

    SciTech Connect

    Not Available

    1992-08-01

    The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  7. United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    SciTech Connect

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

  8. Evaluation of transuranium isotopes inventory for Candu/ACR standard and SEU spent fuel and the possibility to transmute them

    SciTech Connect

    Ghizdeanu, Elena Nineta; Pavelescu, Alexandru; Balaceanu, Victoria

    2007-07-01

    Available in abstract form only. Full text of publication follows: The main disadvantage of nuclear energy is the quantity of long lived radioactive waste produced in a NPP. Transmutation could be one of the solutions to reduce it. Waste transmutation will require a suitable deployment of techniques for spent fuel reprocessing. At present, reprocessing is done by aqueous methods that are very efficient for Pu separation (up to 99.9%). For transmutation applications, new partitioning processes must be developed for minor actinides separation from the high level waste. Although these processes are still very much at the research stage, industrial scale-up will result in the deployment of new, more specific separation techniques for transmutation applications. Partitioning and Transmutation (P and T) techniques could contribute to reduce the radioactive inventory and its associated radio-toxicity. Scientists are looking for ways to drastically reduce both the mass and the radio-toxicity of the nuclear waste to be stored in a deep geological repository, and to reduce the time needed to reach the radioactivity level of the raw material originally used to produce energy. The first stage in the transmutation process is the isotopes inventory formed in the spent fuel. In this paper is made an intercomparison evaluation using WIMS 5B.12 and ORIGEN computer codes. Using these two codes, there is evaluated the isotopes released by a fuel standard from a Candu reactor. Moreover, there is simulated an inventory released by a Candu-SEU reactor and an ACR reactor. (authors)

  9. Passive and Active Radiation Measurements Capability at the INL Zero Power Physics Reactor (ZPPR) Facility

    SciTech Connect

    Robert Neibert; John Zabriskie; Collin Knight; James L. Jones

    2010-12-01

    The Zero Power Physics Reactor (ZPPR) facility is a Department of Energy facility located in the Idaho National Laboratory’s (INL) Materials and Fuels Complex. It contains various nuclear and non-nuclear materials that are available to support many radiation measurement assessments. User-selected, single material, nuclear and non-nuclear materials can be readily utilized with ZPPR clamshell containers with almost no criticality concerns. If custom, multi-material configurations are desired, the ZPPR clamshell or an approved aluminum Inspection Object (IO) Box container may be utilized, yet each specific material configuration will require a criticality assessment. As an example of the specialized material configurations possible, the National Nuclear Security Agency’s Office of Nuclear Verification (NNSA/NA 243) has sponsored the assembly of six material configurations. These are shown in the Appendixes and have been designated for semi-permanent storage that can be available to support various radiation measurement applications.

  10. Preliminary analysis of hot spot factors in an advanced reactor for space electric power systems

    NASA Technical Reports Server (NTRS)

    Lustig, P. H.; Holms, A. G.; Davison, H. W.

    1973-01-01

    The maximum fuel pin temperature for nominal operation in an advanced power reactor is 1370 K. Because of possible nitrogen embrittlement of the clad, the fuel temperature was limited to 1622 K. Assuming simultaneous occurrence of the most adverse conditions a deterministic analysis gave a maximum fuel temperature of 1610 K. A statistical analysis, using a synthesized estimate of the standard deviation for the highest fuel pin temperature, showed probabilities of 0.015 of that pin exceeding the temperature limit by the distribution free Chebyshev inequality and virtually nil assuming a normal distribution. The latter assumption gives a 1463 K maximum temperature at 3 standard deviations, the usually assumed cutoff. Further, the distribution and standard deviation of the fuel-clad gap are the most significant contributions to the uncertainty in the fuel temperature.

  11. Activation characteristics of a solid breeder blanket for a fusion power demonstration reactor

    NASA Astrophysics Data System (ADS)

    Fischer, Ulrich; Tsige-Tamirat, Haileyesus

    2002-12-01

    Activation characteristics have been assessed for a helium cooled solid breeder blanket on the basis of three-dimensional activation calculations for a 2200 MW fusion power demonstration reactor. FISPACT inventory calculations were performed for the beryllium neutron multiplier, the Li 4SiO 4 breeder ceramics and the Eurofer low activation steel. Neutron flux spectra distributions were provided by a previous MCNP calculation. Detailed spatial distributions have been obtained for the nuclide inventories and related quantities such as activity, decay heat and contact dose rate. These data are available form the authors upon request. On the basis of the calculated contact gamma dose rates, the waste quality was assessed with regard to a possible re-use of the activated materials following the remote or the hands-on handling recycling options.

  12. LWR (Light Water Reactor) power plant simulations using the AD10 and AD100 systems

    SciTech Connect

    Wulff, W.; Cheng, H.S.; Chien, C.J.; Jang, J.Y.; Lin, H.C.; Mallen, A.N.; Wang, S.J.; Institute of Nuclear Energy Research, Lung-Tan; Tawian Power Co., Taipei; Brookhaven National Lab., Upton, NY; Institute of Nuclear Energy Research, Lung-Tan )

    1989-01-01

    Boiling (BWR) and Pressurized (PWR) Water Reactor Power Plants are being simulated at BNL with the AD10 and AD100 Peripheral Processor Systems. The AD10 system has been used for BWR simulations since 1984 for safety analyses, emergency training and optimization studies. BWR simulation capabilities have been implemented recently on the AD100 system and PWR simulation capabilities are currently being developed under the auspices of international cooperation. Modeling and simulation methods are presented with emphasis on the simulation of the Nuclear Steam Supply System. Results are presented for BWR simulation and performance characteristics are compared of the AD10 and AD100 systems. It will be shown that the AD100 simulates two times faster than two AD10 processors operating in parallel and that the computing capacity of one AD100 (with FMU processor) is twice as large as that of two AD10 processors. 9 refs., 5 figs., 1 tab.

  13. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    SciTech Connect

    Woo, H.H.; Lu, S.C.

    1981-09-15

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

  14. Evaluation of a Business Case for Safeguards by Design in Nuclear Power Reactors

    SciTech Connect

    Wood, Thomas W.; Seward, Amy M.; Lewis, Valerie A.; Gitau, Ernest TN; Zentner, Michael D.

    2012-12-01

    Safeguards by Design (SbD) is a well-known paradigm for consideration and incorporation of safeguards approaches and associated design features early in the nuclear facility development process. This paradigm has been developed as part of the Next Generation Safeguards Initiative (NGSI), and has been accepted as beneficial in many discussions and papers on NGSI or specific technologies under development within NGSI. The Office of Nuclear Safeguards and Security funded the Pacific Northwest National Laboratory to examine the business case justification of SbD for nuclear power reactors. Ultimately, the implementation of SbD will rely on the designers of nuclear facilities. Therefore, it is important to assess the incentives which will lead designers to adopt SbD as a standard practice for nuclear facility design. This report details the extent to which designers will have compelling economic incentives to adopt SbD.

  15. Survey of electronics capability for SP-100 space reactor power system applications

    NASA Astrophysics Data System (ADS)

    Manvi, Ram; Fujita, Tosh

    Because of reports indicating improvements in the radiation tolerance of some electronic parts, a survey was recently performed by SP-100 project personnel to determine the advisability of revising SP-100 SRPS (space reactor power systems) allowable neutron and gamma dose rates in order to reduce the size and mass of the radiation shield and thereby achieve system mass reductions. The survey results indicate that recent developments to increase the radiation tolerance of a limited set of electronics parts do not justify increasing the allowable SP-100 dose rates for electronic components. Specifically, the recent improvements on a limited set of parts do not justify increasing the current SP-100 allowable specifications of 5 x 10 exp 5 rads gamma dose and 1 x 10 exp 13 neutrons/sq cm fluence. However, if the improvements of 108 rads for gammas and 10 exp 15 neutrons/sq cm can be extended to a wide range of parts, significant mass savings would result.

  16. Survey of electronics capability for SP-100 space reactor power system applications

    NASA Technical Reports Server (NTRS)

    Manvi, Ram; Fujita, Tosh

    1991-01-01

    Because of reports indicating improvements in the radiation tolerance of some electronic parts, a survey was recently performed by SP-100 project personnel to determine the advisability of revising SP-100 SRPS (space reactor power systems) allowable neutron and gamma dose rates in order to reduce the size and mass of the radiation shield and thereby achieve system mass reductions. The survey results indicate that recent developments to increase the radiation tolerance of a limited set of electronics parts do not justify increasing the allowable SP-100 dose rates for electronic components. Specifically, the recent improvements on a limited set of parts do not justify increasing the current SP-100 allowable specifications of 5 x 10 exp 5 rads gamma dose and 1 x 10 exp 13 neutrons/sq cm fluence. However, if the improvements of 108 rads for gammas and 10 exp 15 neutrons/sq cm can be extended to a wide range of parts, significant mass savings would result.

  17. An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing

    SciTech Connect

    Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T.

    1996-05-01

    The objective of this project was to provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel (RPV) could be measured during an annealing process. This project was a coordinated effort between DOE`s Office of Nuclear Energy, Science and Technology; DOE`s Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute`s Non- Destructive Evaluation Center. Ball- thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metal- sheathed thermocouple probes, and 1778 air- suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as shunting (electrical breakdown of insulation separating the thermocouple wires). Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454{degrees}C [850`F], all sensors measured the same temperature within about {plus_minus}5% (23.6{degrees}C [42.5{degrees}F]). Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball- probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball- probe temperature. The model does not predict the temperature as well for the spring-loaded and air suspended probes.

  18. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

    NASA Astrophysics Data System (ADS)

    Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

    2016-01-01

    Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

  19. Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

    NASA Astrophysics Data System (ADS)

    Shoushtari, M. K.; Kakavand, T.; Sadat Kiai, S. M.; Ghaforian, H.

    2010-03-01

    The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity ( ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.

  20. In situ gamma spectrometry of piping in a CANDU heat transport system -- Application during decontamination

    SciTech Connect

    Husain, A.; Breckenridge, C.E.; Storey, D.

    1995-02-01

    An in situ pipe gamma spectrometry technique was applied to determine the activity within piping during various stages of CANDU reactor decontaminations. Measurements were performed in general radiation fields up to {approximately}500 mR/h and required both the detector and the pipe being scanned to be appropriately shielded from other neighboring piping. Measured counts were interpreted using a pipe source efficiency calibration with due regard to its distance dependence. Cobalt-60 was the dominant radionuclide on the piping before the decontamination. Deposition of {sup 124}Sb occurred on out-core piping surfaces during the decontamination. The spectrometry measurements were supplemented with contact radiation field measurements, which were performed using survey detectors housed within specially designed pipe shields. Radiation fields estimated from measured radionuclide activities were compared with the measured radiation fields. On average, the ratio of measured to estimated fields was {approximately}72%. Reasons for this discrepancy are discussed.