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Non-isotropic thermal behavior of an MHTGR fuel block: Impact upon reactivity feedback

Description: This report discusses lumped parameter and detailed multi-dimensional thermal analyses of a New Production Modular High Temperature Gas Cooled Reactor (NP-MHTGR) fuel block were conducted that indicated that during a power transient, the target temperature would rise significantly later than the fuel temperature. This behavior, which is due to radiative, convective and conductive heat transport phenomena within the fuel block coupled with the significantly different thermal physical properties of the fuel block materials, leads to the potential for a delayed positive contribution to the temperature coefficient of reactivity of the NP-MHTGR core during TOP events. These results have indicated the need for additional experimental and analytical studies in order to more fully assess the design, operational and safety implications of this phenomenon. In addition, experiments in the TREAT facility are planned to provide additional data to assist in the capabilities to predict the reactivity feedback characteristics of the NP-MHTGR core. These studies will be the subject of a future paper.
Date: September 1, 1992
Creator: Delpech, M.; Singer, R. M. & Finck, P. J.
Partner: UNT Libraries Government Documents Department

Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

Description: A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.
Date: April 1, 1994
Creator: Forsberg, C. W.; Conklin, J. & Reich, W. J.
Partner: UNT Libraries Government Documents Department

Characteristics of potential repository wastes. Volume 2

Description: The LWR spent fuels discussed in Volume 1 of this report comprise about 99% of all domestic non-reprocessed spent fuel. In this report we discuss other types of spent fuels which, although small in relative quantity, consist of a number of diverse types, sizes, and compositions. Many of these fuels are candidates for repository disposal. Some non-LWR spent fuels are currently reprocessed or are scheduled for reprocessing in DOE facilities at the Savannah River Site, Hanford Site, and the Idaho National Engineering Laboratory. It appears likely that the reprocessing of fuels that have been reprocessed in the past will continue and that the resulting high-level wastes will become part of defense HLW. However, it is not entirely clear in some cases whether a given fuel will be reprocessed, especially in cases where pretreatment may be needed before reprocessing, or where the enrichment is not high enough to make reprocessing attractive. Some fuels may be canistered, while others may require special means of disposal. The major categories covered in this chapter include HTGR spent fuel from the Fort St. Vrain and Peach Bottom-1 reactors, research and test reactor fuels, and miscellaneous fuels, and wastes generated from the decommissioning of facilities.
Date: July 1, 1992
Partner: UNT Libraries Government Documents Department

Temperature-Initiated Passive Cooling System (TIPACS)

Description: The Temperature-Initiated Passive Cooling System (TIPACS) is a recently invented passive cooling system that transfers heat from a hot, insulated system to a cooler, external environment. TIPACS has four defining characteristics: efficient heat-transfer, passive with no moving components, thermal switch mechanism that allows heat transfer only above a preset temperature, and one-way (heat diode) heat transfer. Example applications include cooling (1) building attics, (2) electrical sheds, (3) chemical reactors, (4) utility-load-leveling batteries, and (5) nuclear reactor containments. TIPACS was evaluated for cooling a modular high-temperature gas-cooled reactor (MHTGR) cavity. This evaluation indicates potential performance and economic advantages.
Date: May 16, 1994
Creator: Forsberg, C. W. & Conklin, J. C.
Partner: UNT Libraries Government Documents Department

The gas turbine-modular helium reactor (GT-MHR), high efficiency, cost competitive, nuclear energy for the next century

Description: The Gas Turbine-Modular Helium Reactor (GT-MHR) is the result of coupling the evolution of a small passively safe reactor with key technology developments in the US during the last decade: large industrial gas turbines, large active magnetic bearings, and compact, highly effective plate-fin heat exchangers. The GT-MHR is the only reactor concept which provides a step increase in economic performance combined with increased safety. This is accomplished through its unique utilization of the Brayton cycle to produce electricity directly with the high temperature helium primary coolant from the reactor directly driving the gas turbine electrical generator. This cannot be accomplished with another reactor concept. It retains the high levels of passive safety and the standardized modular design of the steam cycle MHTGR, while showing promise for a significant reduction in power generating costs by increasing plant net efficiency to a remarkable 47%.
Date: April 1, 1994
Creator: Zgliczynski, J. B.; Silady, F. A. & Neylan, A. J.
Partner: UNT Libraries Government Documents Department

Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors

Description: During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs) the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS), are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This paper uses the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits.
Date: September 1, 1992
Creator: Kroeger, P. G.
Partner: UNT Libraries Government Documents Department

Evaluation of MHTGR fuel reliability

Description: Modular High-Temperature Gas-Cooled Reactor (MHTGR) concepts that house the reactor vessel in a tight but unsealed reactor building place heightened importance on the reliability of the fuel particle coatings as fission product barriers. Though accident consequence analyses continue to show favorable results, the increased dependence on one type of barrier, in addition to a number of other factors, has caused the Nuclear Regulatory Commission (NRC) to consider conservative assumptions regarding fuel behavior. For this purpose, the concept termed ``weak fuel`` has been proposed on an interim basis. ``Weak fuel`` is a penalty imposed on consequence analyses whereby the fuel is assumed to respond less favorably to environmental conditions than predicted by behavioral models. The rationale for adopting this penalty, as well as conditions that would permit its reduction or elimination, are examined in this report. The evaluation includes an examination of possible fuel-manufacturing defects, quality-control procedures for defect detection, and the mechanisms by which fuel defects may lead to failure.
Date: July 1, 1992
Creator: Wichner, R. P. & Barthold, W. P.
Partner: UNT Libraries Government Documents Department

MHTGR thermal performance envelopes: Reliability by design

Description: This document discusses thermal performance envelopes which are used to specify steady-state design requirements for the systems of the Modular High Temperature Gas-Cooled Reactor to maximize plant performance reliability with optimized design. The thermal performance envelopes are constructed around the expected operating point accounting for uncertainties in actual plant as-built parameters and plant operation. The components are then designed to perform successfully at all points within the envelope. As a result, plant reliability is maximized by accounting for component thermal performance variation in the design. The design is optimized by providing a means to determine required margins in a disciplined and visible fashion.
Date: May 1, 1992
Creator: Etzel, K. T.; Howard, W. W. & Zgliczynski, J. B.
Partner: UNT Libraries Government Documents Department

A vented low pressure containment strategy for the Modular High Temperature Gas-Cooled Reactor (MHTGR)

Description: This paper presents the response of the 450 MW(t) MHTGR with a steam turbine power conversion system to expected and hypothetical accident source term assumptions. A range of vented low pressure containment (VLPC) strategies was considered that would enhance the retention of radionuclides. This study was prepared to review the technical merits of VLPC options in response to an NRC request during preapplication review of the steam cycle MMGR. The study found that, even under arbitrary hypothetical assumptions regarding significantly lower than expected fuel performance, vented low pressure containment options can effectively reduce accident doses. The reference design with a VLPC meets the 10CFR100 and prompt fatality doses even with lower than expected fuel performance. Alternative VLPC designs were studied which could be used to augment the current design to provide additional margin.
Date: April 1, 1994
Creator: Dilling, D.; Dunn, T. D. & Silady, F. A.
Partner: UNT Libraries Government Documents Department

Impact of increasing MHTGR power on passive heat removal

Description: In 1990 a cost reduction study recommended that the reference US MHTGR module design be changed to an 84-column, 450 MW(t) annular reactor core to attain improved economics with the same high level of safety as the previous reference 66-column, 350 MW(t) MHTGR module. The objective of this paper is to report on a recently completed core configuration trade study that reviewed the basis for that recommendation with more detailed assessments. The trade study examined alternate core configurations in terms of the size, shape, and power level. Core configurations at 450 MW(t), an alterative at higher power, and one at lower power were considered. These alternatives represented the maximum achievable power for fuel element for two different reactor vessel sizes. Fuel, reactor internal and vessel temperatures during pressurized and depressurized conduction cooldown transients are presented and compared to limits. Based on the need to improve economics without sacrificing the MHTGR`s high level of safety, the trade study confirmed that the previously selected 84-column, 450 MW(t) annular design remains the preferable configuration.
Date: July 1, 1992
Creator: Dunn, T. D.; Schwartz, A. A. & Silady, F. A.
Partner: UNT Libraries Government Documents Department

Production test IP-376-D, Supplement B Irradiation of MGCR-HDR-3 Test Element

Description: The objective of this supplement to PT-IP-376-D, Irradiation of MGCR-HDR-3 Test Element is to authorize 1000 hours of operation at a maximum test specimen surface temperature of 1700 F. The original production test authorized a test duration of four months at a maximum specimen surface temperature of 1500 F; supplement A authorized extension of the test duration to ten months. The desired increase in surface temperature is requested to demonstrate the general feasibility of operation of the fuel element at 1700 F, and to obtain specific information on the performance of Hastelloy-X cladding and fuel bodies. The increased temperature has been approved by the Atomic Energy Commission.
Date: July 11, 1961
Creator: Baars, R. E.
Partner: UNT Libraries Government Documents Department

Nuclear reactors built, being built, or planned, 1991

Description: This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).
Date: July 1, 1992
Creator: Simpson, B.
Partner: UNT Libraries Government Documents Department

Continuous improvement of the MHTGR safety and competitive performance

Description: An increase in reactor module power from 350 to 450 MW(t) would markedly improve the economics of the Modular High Temperature Gas-Cooled Reactor (MHTGR). The higher power level was recommended as the result of an in-depth cost reduction study undertaken to compete with the declining price of fossil fuel. The safety assessment confirms that the high level of safety, which relies on inherent characteristics and passive features, is maintained at the elevated power level. Preliminary systems, nuclear, and safety performance results are discussed for the recommended 450 MW(t) design. Optimization of plant parameters and design modifications accommodated the operation of the steam generator and circulator at the higher power level. Events in which forced cooling is lost, designated as conduction cooldowns are described in detail. For the depressurized conduction cooldown, without full helium inventory, peak fuel temperatures are significantly lowered. A more negative temperature coefficient of reactivity was achieved while maintaining an adequate fuel cycle and reactivity control. Continual improvement of the MHTGR delivers competitive performance without relinquishing the high safety margins demanded of the next generation of power plants.
Date: May 1, 1992
Creator: Eichenberg, T. W.; Etzel, K. T.; Mascaro, L. L. & Rucker, R. A.
Partner: UNT Libraries Government Documents Department

Nuclear Safety. Technical Progress Journal, October--December 1991: Volume 32, No. 4

Description: This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.
Date: January 1, 1991
Partner: UNT Libraries Government Documents Department

A review of selected aspects of the effect of water vapor on fission gas release from uranium oxycarbide

Description: A selective review is presented of previous measurements and the analysis of experiments on the effect of water vapor on fission gas release from uranium oxycarbide. Evidence for the time-dependent composition of the uranium oxycarbide fuel; the diffusional release of fission gas; and the initial, rapid and limited release of stored fission gas is discussed. In regard to the initial, rapid release of fission gas, clear restrictions on mechanistic hypotheses can be deduced from the experimental data. However, more fundamental experiments may be required to establish the mechanism of the rapid release.
Date: April 1, 1994
Creator: Myers, B. F.
Partner: UNT Libraries Government Documents Department

Applications of Monte Carlo methods for the analysis of MHTGR case of the PROTEUS benchmark

Description: Monte Carlo methods, as implemented in the MCNP code, have been used to analyze the neutronics characteristics of benchmarks related to Modular High Temperature Gas-Cooled Reactors. The benchmarks are idealized versions of the Japanes (VHTRC) and Swiss (PROTEUS) facilities and an actual configurations of the PROTEUS Configuration I experiment. The purpose of the unit cell benchmarks is to compare multiplication constants, critical bucklings, migration lengths, reaction rates and spectral indices. The purpose of the full reactors benchmarks is to compare multiplication constants, reaction rates, spectral indices, neutron balances, reaction rates profiles, temperature coefficients of reactivity and effective delayed neutron fractions. All of these parameters can be calculated by MCNP, which can provide a very detailed model of the geometry of the configurations, from fuel particles to entire fuel assemblies, using at the same time a continuous energy model. These characteristics make MCNP a very useful tool to analyze these MHTGR benchmarks. We have used the MCNP latest version, 4.x, eld = 01/12/93 with an ENDF/B-V cross section library. This library does not yet contain temperature dependent resonance materials, so all calculations correspond to room temperature, T = 300{degree}K. Two separate reports were made -- one for the VHTRC, the other for the PROTEUS benchmark.
Date: April 1, 1994
Creator: Difilippo, F. C.
Partner: UNT Libraries Government Documents Department

Applications of Monte Carlo methods for the analysis of MHTGR case of the VHTRC benchmark

Description: Monte Carlo methods, as implemented in the MCNP code, have been used to analyze the neutronics characteristics of benchmarks related to Modular High Temperature Gas-Cooled Reactors. The benchmarks are idealized versions of the Japanese (VHTRC) and Swiss (PROTEUS) facilities and an actual configuration of the PROTEUS Configuration 1 experiment. The purpose of the unit cell benchmarks is to compare multiplication constants, critical bucklings, migration lengths, reaction rates and spectral indices. The purpose of the full reactors benchmarks is to compare multiplication constants, reaction rates, spectral indices, neutron balances, reaction rates profiles, temperature coefficients of reactivity and effective delayed neutron fractions. All of these parameters can be calculated by MCNP, which can provide a very detailed model of the geometry of the configurations, from fuel particles to entire fuel assemblies, using at the same time a continuous energy model. These characteristics make MCNP a very useful tool to analyze these MHTGR benchmarks. The author has used the MCNP latest version, 4.x, eld = 01/12/93 with an ENDF/B-V cross section library. This library does not yet contain temperature dependent resonance materials, so all calculations correspond to room temperature, T = 300{degrees}K. Two separate reports were made -- one for the VHTRC, the other for the PROTEUS benchmark.
Date: March 1, 1994
Creator: Difilippo, F. C.
Partner: UNT Libraries Government Documents Department

Interactive simulations of gas-turbine modular HTGR transients and heatup accidents

Description: An interactive workstation-based simulator has been developed for performing analyses of modular high-temperature gas-cooled reactor (MHTGR) core transients and accidents. It was originally developed at Oak Ridge National Laboratory for the US Nuclear Regulatory Commission to assess the licensability of the US Department of Energy (DOE) steam cycle design 350-MW(t) MHTGR. Subsequently, the code was modified under DOE sponsorship to simulate the 450-MW(t) Gas Turbine (GT) design and to aid in development and design studies. Features of the code (MORECA-GT) include detailed modeling of 3-D core thermal-hydraulics, interactive workstation capabilities that allow user/analyst or ``operator`` involvement in accident scenarios, and options for studying anticipated transients without scram (ATWS) events. In addition to the detailed models for the core, MORECA includes models for the vessel, Shutdown Cooling System (SCS), and Reactor Cavity Cooling System (RCCS), and core point kinetics to accommodate ATWS events. The balance of plant (BOP) is currently not modeled. The interactive workstation features include options for on-line parameter plots and 3-D graphic temperature profiling. The studies to date show that the proposed MHTGR designs are very robust and can generally withstand the consequences of even the extremely low probability postulated accidents with little or no damage to the reactor`s fuel or metallic components.
Date: June 1, 1994
Creator: Ball, S. J. & Nypaver, D. J.
Partner: UNT Libraries Government Documents Department

Innovative safety features of the modular HTGR

Description: The Modular High Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry, and the utilities. Near-term development is focused on electricity generation. The top-level safety requirement is that the plant`s operation not disturb the normal day-to-day activities of the public. Quantitatively, this requires that the design meet the US Environmental Protection Agency`s Protective Action Guides at the site boundary and hence preclude the need for sheltering or evacuation of the public. To meet these stringent safety requirements and at the same time provide a cost competitive design requires the innovative use of the basic high temperature gas-cooled reactor features of ceramic fuel, helium coolant, and a graphite moderator. The specific fuel composition and core size and configuration have been selected to the use the natural characteristics of these materials to develop significantly higher margins of safety. In this document the innovative safety features of the MHTGR are reviewed by examining the safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles. A broad range of challenges to core heat removal are examined, including a loss of helium pressure of a simultaneous loss of forced cooling of the core. The challenges to control of heat generation consider not only the failure to insert the reactivity control systems but also the withdrawal of control rods. Finally, challenges to control of chemical attack of the ceramic-coated fuel are considered, including catastrophic failure of the steam generator, which allows water ingress, or failure of the pressure vessels, which allows air ingress. The plant`s response to these extreme challenges is not dependent on operator action, and the events considered encompass conceivable operator errors.
Date: January 1, 1992
Creator: Silady, F. A. & Simon, W. A.
Partner: UNT Libraries Government Documents Department

Innovative safety features of the modular HTGR. Revision 1

Description: In this document the innovative safety features of the MHTGR are reviewed by examining the safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles. A broad range of challenges to core heat removal are examined, including a loss of helium pressure and a simultaneous loss of forced cool of the core.
Date: April 1, 1992
Creator: Silady, F. A. & Simon, W. A.
Partner: UNT Libraries Government Documents Department

Recovery of weapon plutonium as feed material for reactor fuel

Description: This report presents preliminary considerations for recovering and converting weapon plutonium from various US weapon forms into feed material for fabrication of reactor fuel elements. An ongoing DOE study addresses the disposition of excess weapon plutonium through its use as fuel for nuclear power reactors and subsequent disposal as spent fuel. The spent fuel would have characteristics similar to those of commercial power spent fuel and could be similarly disposed of in a geologic repository.
Date: March 16, 1994
Creator: Armantrout, G. A.; Bronson, M. A. & Choi, Jor-Shan
Partner: UNT Libraries Government Documents Department

Options for treating high-temperature gas-cooled reactor fuel for repository disposal

Description: This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.
Date: February 1, 1992
Creator: Lotts, A. L.; Bond, W. D.; Forsberg, C. W.; Glass, R. W.; Harrington, F. E.; Micheals, G. E. et al.
Partner: UNT Libraries Government Documents Department

Production test IP-376-D: Irradiation of MGCR-HDR-3 test element

Description: The objective of this test detailed in this report is to irradiate a test fuel assembly for the MGCR (Maritime Gas Cooled Reactor). The irradiation of this assembly will be carried out in the DR-1 Loop under controlled conditions to determine the feasibility of the heterogeneous 19-rod bundle fuel element concept. Diffusion of fission products through metal cladding. Fission gas retention of fuel bodies. Dimensional stability of fuel bodies, and satisfactory performance of the creep-shrink process for maintaining pellet position in the fuel pin.
Date: November 29, 1960
Creator: Bennett, E. C.
Partner: UNT Libraries Government Documents Department