25 Matching Results

Search Results

Advanced search parameters have been applied.

MHTGR-Nuclear Island Engineering: Final summary report for the period November 30, 1987 through December 1, 1988

Description: This report summarizes the Modular High-Temperature Gas-Cooled Reactor (MHTGR) - Nuclear Island Engineering (NIE) design and development work performed by General Atomics (GA) for the period November 30, 1987 through December 1, 1988, under the Department of Energy (DOE) Contract AC03-88SF17367. The scope of the report includes work performed by Bechtel National Inc. (BNI), Combustion Engineering Inc. (C-E), and James Howden Company, as major subcontractors to GA.
Date: December 1, 1988
Partner: UNT Libraries Government Documents Department

Fission product plateout/liftoff/washoff test plan. Revision 1

Description: A test program is planned in the COMEDIE loop of the Commissariat a l`Energy Atomique (CEA), Grenoble, France, to generate integral test data for the validation of computer codes used to predict fission product transport and core corrosion in the Modular High Temperature Gas-Cooled Reactor (MHTGR). The inpile testing will be performed by the CEA under contract from the US Department of Energy (DOE); the contract will be administered by Oak Ridge National Laboratory (ORNL). The primary purpose of this test plan is to provide an overview of the proposed program in terms of the overall scope and schedule. 8 refs, 3 figs.
Date: May 1, 1988
Creator: Acharya, R. & Hanson, D.
Partner: UNT Libraries Government Documents Department

Predicted discharge plutonium isotopics for LEU [low-enriched uranium] test pebble irradiated in the AVR [Arbeitsgemeinschaft Versuchsreaktor]

Description: A Subprogram Plan related to the Arbeitsgemeinschaft Versuchsreaktor (AVR) Test Program is in place and describes cooperative work being carried out under the United States/Federal Republic of Germany (US/FRG) Implementing Agreement for Cooperation in Gas-Cooled Reactor Development. The AVR information to be provided as described in the plan will provide a basis for examining the accuracy of computational methods used for performance and safety analysis. The purpose of the cooperation is to obtain experimental information from the AVR relevant to the performance and safety of modular gas-cooled reactors, and to compare measured results with predictions of analytical tools. This report provides a progress report on the prediction of plutonium buildup in LEU fuel in a high-enriched uranium (HEU) core and also describes the method for calculating the U-238 resonance integral (cross section). 4 refs., 5 figs., 11 tabs.
Date: June 1, 1988
Creator: Lane, R.K. & Lefler, W.L.
Partner: UNT Libraries Government Documents Department

Metals design handbook

Description: This report gives an approved set of material properties over a range of environmental conditions which are sufficient to design the metallic components in the reactor system and hot duct assembly. Table 1-1 list these metallic components together with the reference design material chosen for each component. Table 1-2 summarizes the structural criteria of each metallic component taken from the component specifications. In all cases, the criteria references the ASME B&PV Code. The ASME-Code includes the material properties of Coded material. The Code does not, however, include environmental effects (such as irradiation, corrosion, or thermal aging), and for some components the material maximum allowable temperature is below that of the design and/or postulated ``safety-related`` accident conditions. Table 1-3 gives the Code limits for the portions of the Code given in Table 1-2. This document includes the effects of the radiation environment, chemical impurity effects (in the primary coolant), and the effects of thermal aging and corrosion on the metallic properties. The design information introduced in this document includes that available from the ASME B&PV Code High-Temperature Code Cases plus material information from General Atomics (GA) and Oak Ridge National Laboratories (ORNL) that is published.
Date: July 1, 1988
Creator: Betts, W.S.
Partner: UNT Libraries Government Documents Department

Development of improved TRISO-P fuel particle P-PyC coating

Description: Low defect fuels are required for the MHTGR to meet tighter fuel performance for this reactor design (Ref. 1). Exposed heavy metal (HM) contamination levels must be reduced to {le} 1E-5 fraction. Particle coating breakage during the fuel compact fabrication process has been shown to be a major source of HM contamination in the final fuel compacts. Excessive forces are experienced by the coated fuel particles during matrix injection, which leads to coating failure. Adding a sacrificial, low Young`s modulus, overcoating of low density PyC in a fluidized particle bed, was shown to greatly increase the crush strength of TRISO coated fuel particles in 1986 studies (Ref. 2). The new TRISO coated fuel particle design was designated the TRISO-P coated fuel particle type. In 1987, the TRISO-P particle type was used to produce low defect fuel compacts for irradiation in the HRB-21 Capsule (Ref. 3). However, the exposed HM contamination levels for that fuel barely met the product specification limit of {le} 1.0E-5. The small margin of safety between product quality and the specification limit dictated that additional process development of the TRISO-P particle design must be conducted. This document discusses the program scope, requirements, documentation and schedule.
Date: April 29, 1988
Creator: Adams, C.C.
Partner: UNT Libraries Government Documents Department

Graphite design handbook

Description: The objectives of the Graphite Design Handbook (GDH) are to provide and maintain a single source of graphite properties and phenomenological model of mechanical behavior to be used for design of MHTGR graphite components of the Reactor System, namely, core support, permanent side reflector, hexagonal reflector elements, and prismatic fuel elements; to provide a single source of data and material models for use in MHTGR graphite component design, performance, and safety analyses; to present properties and equations representing material models in a form which can be directly used by the designer or analyst without the need for interpretation and is compatible with analytical methods and structural criteria used in the MHTGR project, and to control the properties and material models used in the MHTGR design and analysis to proper Quality Assurance standards and project requirements. The reference graphite in the reactor internal components is the nuclear grade 2020. There are two subgrades of interest, the cylinder nuclear grade and the large rectangular nuclear grade. The large rectangular nuclear grade is molded in large rectangular blocks. It is the reference material for the permanent side reflector and the central column support structure. The cylindrical nuclear grade is isostatically pressed and is intended for use as the core support component. This report gives the design properties for both H-451 and 2020 graphite as they apply to their respective criteria. The properties are presented in a form for design, performance, and safety calculations that define or validate the component design. 103 refs., 20 figs., 19 tabs.
Date: September 1, 1988
Creator: Ho, F.H.
Partner: UNT Libraries Government Documents Department

Tritium distribution in the MHTGR

Description: The {sup 3}H production, transport and environmental release from the 350 MW(t) Modular High Temperature Gas Cooled Reactor was analyzed. The analysis was performed using a modified TRITGO computer code, plant data base from the Preliminary Safety Information Document and materials property data from the Fuel Design Data Manual, Issue F. The analysis indicates that most of the {sup 3}H produced in the reactor is retained by the fuel particles and the structural graphite elements. The single largest source of {sup 3}H is ternary fission in the fuel particles, of which 95% is retained by the particles. The {sup 3}H released from the core and the {sup 3}H produced by {sup 3}He activation are largely removed by the Helium Purification System. Assuming zero leakage of water from the secondary system, the average predicted {sup 3}H activity in the secondary water of 0.35 {mu}Ci/g is much greater than the allowable activity of 5 pCi/g for direct discharge into the environment. If any of the secondary water has to be discharged, it must be diluted prior to discharge. 10 refs., 9 figs.
Date: May 1, 1988
Creator: Acharya, R.
Partner: UNT Libraries Government Documents Department

H-451 graphite irradiation creep design model; Revision 1

Description: Available irradiation creep data on H-451 graphite area analyzed and fitted to the proposed creep model in a standard linear solid (a linear viscoelastic model). A creep equation is obtained and recommended for preliminary design use. It is found that the regression is significant and the creep equation is a good predictor. The standard error (SE) of the estimate is smaller than that used in the core graphite criteria development. This smaller SE shall be used in all future work related to criteria development. The creep coefficient and/or model can be further improved if additional creep data can be obtained. For this purpose several creep experiments are recommended. The immediate one is to capsule 87M-2A currently under design.
Date: July 1, 1988
Partner: UNT Libraries Government Documents Department

MHTGR [modular high-temperature gas-cooled reactor] core physics validation plan

Description: This document contains the verification and validation (V&V) plan for analytical methods utilized in the nuclear design for normal and off-normal conditions within the Modular High-Temperature Gas-Cooled Reactor (MHTGR). Regulations, regulatory guides, and industry standards have been reviewed and the approach for V&V has been developed. MHTGR core physics methods are described and the status of previous V&V is summarized within this document. Additional work required to verify and validate these methods is identified. The additional validation work includes comparison of calculations with available experimental data, benchmark comparison of calculations with available experimental data, benchmark comparisons with other validated codes, results from a cooperative program now underway at the Arbeitsgemeinschaft Versuchs-Reaktor GmbH (AVR) facility in Germany, results from a planned series of experiments on the Compact Nuclear Power Source (CNPS) facility at Los Alamos, and detailed documentation of all V&V studies. In addition, information will be obtained from planned international cooperative agreements to provide supplemental data for V&V. The regulatory technology development plan will be revised to include these additional experiments. A work schedule and cost estimate for completing this plan is also provided. This work schedule indicates the timeframe in which major milestones must be performed in order to complete V&V tasks prior to the issuance of preliminary design approval from the NRC. The cost to complete V&V tasks for core physics computational methods is estimated to be $2.2M. 41 refs., 13 figs., 8 tabs.
Date: January 1, 1988
Creator: Baxter, A. & Hackney, R.
Partner: UNT Libraries Government Documents Department

Multiaxial graphite test specimen

Description: A multiaxial test program is to be conducted by Oak Ridge National Laboratory (ORNL) on the core component graphite. The objectives of the tests are to obtain failure data under uniaxial and biaxial states of stress in order to construct a failure surface in a two-dimensional stress space. These data will be used in verifying the accuracy of the maximum stress failure theory being proposed for use in designing the core graphite components. Tubular specimens are proposed to be used and are either loaded axially and/or subjected to internal pressure. This report includes a study on three specimen configurations. The conclusions of that study indicate that an elliptical transition geometry procedures the smallest discontinuity effects. Several loading combustions were studied using the elliptical transition specimen. The primary purpose is to establish the location of the highest stress state and its relation to the gage section for all of the loading conditions. The tension/internal pres sure loading condition (1:1) indicated that the high stress area is just outside the gage section but still should be acceptable. 5 refs., 18 figs.
Date: September 1, 1988
Partner: UNT Libraries Government Documents Department

MHTGR: New production reactor summary of experience base

Description: Worldwide interest in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) stems from the capability of the system to retain the advanced fuel and thermal performance while providing unparalleled levels of safety. The small power level of the MHTGR and its passive systems give it a margin of safety not attained by other concepts being developed for power generation. This report covers the experience base for the key nuclear system, components, and processes related to the MHTGR-NPR. 9 refs., 39 figs., 9 tabs.
Date: March 1, 1988
Partner: UNT Libraries Government Documents Department

Severe accident core heatup transients in modular high temperature gas-cooled reactors without operating Reactor Cavity Cooling Systems

Description: The ultimate decay heat removal system for the current Modular High Temperature Gas-Cooled reactors is a completely passive natural convection air cooling loop. This paper considers an extremely remote accident scenario, where even this passive system fails, and heat rejection is only via a layer of thermal insulation to the reactor silo structure and the surrounding soil. The results show that even in this case the peak fuel temperatures remain well within safe limits. However, vessel and concrete temperatures can - under extreme circumstances and after several weeks - reach temperature levels at which structural failure becomes possible.
Date: January 1, 1988
Creator: Kroeger, P.G.
Partner: UNT Libraries Government Documents Department

MHTGR (Modular High-Temperature Gas-Cooled Reactor) technology development plan

Description: This paper presents the approach used to define the technology program needed to support design and licensing of a Modular High-Temperature Gas-Cooled Reactor (MHTGR). The MHTGR design depends heavily on data and information developed during the past 25 years to support large HTGR (LHTGR) designs. The technology program focuses on MHTGR-specific operating and accident conditions, and on validation of models and assumptions developed using LHTGR data. The technology program is briefly outlined, and a schedule is presented for completion of technology work which is consistent with completion of a Final Safety Summary Analysis Report (FSSAR) by 1992.
Date: January 1, 1988
Creator: Homan, F.J. & Neylan, A.J.
Partner: UNT Libraries Government Documents Department

The passive safety characteristics of modular high temperature gas-cooled reactor fuel elements

Description: High-Temperature Gas-Cooled Reactors (HTGR) in both the US and West Germany use an all-ceramic, coated fuel particle to retain fission products. Data from irradiation, postirradiation examinations and postirradiation heating experiments are used to study the performance capabilities of the fuel particles. The experimental results from fission product release tests with HTGR fuel are discussed. These data are used for development of predictive fuel performance models for purposes of design, licensing, and risk analyses. During off normal events, where temperatures may reach up to 1600/degree/C, the data show that no significant radionuclide releases from the fuel will occur.
Date: January 1, 1988
Creator: Goodin, D.T.; Kania, M.J.; Nabielek, H.; Schenk, W. & Verfondern, K.
Partner: UNT Libraries Government Documents Department

Hypothetical air ingress scenarios in advanced modular high temperature gas cooled reactors

Description: Considering an extremely hypothetical scenario of complete cross duct failure and unlimited air supply into the reactor vessel of a modular high temperature gas cooled ractor, it is found that the potential air inflow remains limited due to the high friction pressure drop through the active core. All incoming air will be oxidized to CO and some local external burning would be temporarily possible in such a scenario. The accident would have to continue with unlimited air supply for hundreds of hours before the core structural integrity would be jeopardized.
Date: January 1, 1988
Creator: Kroeger, P.G.
Partner: UNT Libraries Government Documents Department

Reactor-specific spent fuel discharge projections, 1987-2020

Description: The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs.
Date: March 1, 1988
Creator: Walling, R.C.; Heeb, C.M. & Purcell, W.L.
Partner: UNT Libraries Government Documents Department

Depressurized core heatup accident scenarios in advanced modular high temperature gas cooled reactors

Description: The decay heat removal by a passive air colling system from a modular high tempeature gas cooled reactor during depressurized core heatup accident scenarios was analyzed. The effects of several design and operating parameters on the peak fuel and vessel temperatures were established. The results indicate that fuel and vessel temperatures remain well below failure levels and that significant safety margins exist in the key variables of decay heat and core thermal conductivities.
Date: January 1, 1988
Creator: Kroeger, P.G.
Partner: UNT Libraries Government Documents Department

Economic characteristics of a smaller, simpler reactor

Description: Reduced load growth and heightened concern with economic risk has led to an expressed utility preference for smaller capacity additions. The Modular High Temperature Reactor (MHTGR) plant has been developed as a small, simple plant that has limited financial risk and is economically competitive with comparatively sized coal plants. Competitive economics is achieved by the simplifications made possible in a small MHTGR, reduction in the quantity of nuclear grade construction and design standardization and certification. Assessments show the MHTGR plant to have an economic advantage over coal plants for plant sizes from 270 MWe to 1080 MWe. Financial risk is limited by small unit sizes and short lead times that allow incremental deployment. Evaluations show the MHTGR incremental deployment capability to reduce negative cash flows by almost a factor of 2 relative to that required by a single large nuclear plant.
Date: January 1, 1988
Creator: LaBar, M. & Bowers, H.
Partner: UNT Libraries Government Documents Department

The licensing experinece of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

Description: The MHTGR is an advanced reactor concept being developed under a cooperative program involving the US Government, the nuclear industry, and the utilities. The design utilizes the basic HTGR features of ceramic fuel, helium coolant, and a graphite moderator. However, the specific size and configuration are selected to utilize the inherent characteristics of these materials to develop passive safety features that provide a significantly higher margin of safety than current generation reactors. The design meets the US Environmental Protection Agency's Protective Action Guidelines at the site boundary, hence precluding the need for sheltering or vacation of the public during any licensing basis event. This safe behavior is not dependent upon operator action and is insensitive to operator error. The MHTGR Licensing Plan agreed to with the US Nuclear Regulatory Commission (NRC) is discussed with particular attention to the framework of the preapplication review. The objective and scope of each key document prepared for the NRC review is presented. A summary is provided of the safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles. The regulatory interaction process and results are discussed through the NRC staff, NRC contractor, and ACRS reviews. 11 refs., 3 figs.
Date: September 1, 1988
Creator: Silady, F.A.; Cunliffe, J.C. & Walker, L.P.
Partner: UNT Libraries Government Documents Department

Environmental aspects of MHTGR (Modular High-Temperature Gas-Cooled Reactor) operation

Description: The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept being developed under a cooperative program involving the US Government, the utilities and the nuclear industry. This plant design utilizes basic High Temperature Gas-Cooled Reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The MHTGR design approach leading to exceptional safety performance also leads to plant operation which is characterized by extremely low radiological emissions even for very low probability accidents. Coated fuel particles retain radionuclides within the fuel, thus minimizing material contamination and personnel exposure. The objective of this paper is to characterize radioactive effluents expected from the normal operation of an MHTGR. In addition, other nonradioactive effluents associated with a power generating facility are discussed. Nuclear power plants produce radioactive effluents during normal operation in gaseous, liquid and solid forms. Principal sources of radioactive waste within the MHTGR are identified. The manner in which it is planned to treat these wastes is described. Like other reactors, the MHTGR produces nonradioactive effluents associated with heat generation and chemical usage. However, due to the MHTGR's higher efficiency, water usage requirements and chemical discharges for the MHTGR are minimized relative to other types of nuclear power plants. Based upon prior operating HTGR experience and analysis, effluents are quantified in terms of radioactivity levels and/or emission volume. Results, quantified within the paper, demonstrate that effluents from the MHTGR are well below regulatory limits and that the MHTGR has a minimal impact upon the public and the environment. 14 refs., 2 figs., 4 tabs.
Date: September 1, 1988
Creator: Neylan, A.J.; Dilling, D.A. & Cardito, J.M.
Partner: UNT Libraries Government Documents Department

MHTGR (Modular High-Temperature Gas-Cooled Reactor) design and development status

Description: The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced power plant concept which has been under design definition since 1984. The design utilizes basic high-temperature gas-cooled reactor features of ceramic fuel, helium coolant and a graphite moderator which have been under development for 30 years. The geometric arrangement of the reactor vessels, the core and the heat removal components has been selected to exploit the inherent characteristics associated with high temperature materials. The design utilizes passively safe features which provide a higher margin of safety and investment protection than current generation reactors. The design has been evaluated to be economically attractive relative to modern coal fired plants. The design and development program is a cooperative effort by the US government, the utilities and the nuclear industry. 8 refs., 4 figs., 4 tabs.
Date: August 1, 1988
Creator: Turner, R.F. & Neylan, A.J.
Partner: UNT Libraries Government Documents Department

Modular High-Temperature Gas-Cooled Reactor short term thermal response to flow and reactivity transients

Description: The analyses reported here have been conducted at the Oak Ridge National Laboratory (ORNL) for the US Nuclear Regulatory Commission's (NRC's) Division of Regulatory Applications of the Office of Nuclear Regulatory Research. The short-term thermal response of the Modular High-Temperature Gas-Cooled Reactor (MHTGR) is analyzed for a range of flow and reactivity transients. These include loss of forced circulation (LOFC) without scram, moisture ingress, spurious withdrawal of a control rod group, hypothetical large and rapid positive reactivity insertion, and a rapid core cooling event. The coupled heat transfer-neutron kinetics model is also described.
Date: January 1, 1988
Creator: Cleveland, J.C.
Partner: UNT Libraries Government Documents Department

Design requirements for high-temperature metallic component materials in the US modular HTGR

Description: The modular high temperature gas-cooled reactor (MHTGR) is a 350 MW(t) second generation reactor system design which during normal operation circulates helium with a mixed mean coal and hot temperature of 260/sup 0/C (500/sup 0/C) and 690/sup 0/C (1270/sup 0/F), respectively. The design incorporates passive design features which allow the plant to be safely shutdown and cooled with no active systems or operator action being required. A key feature of this concept is the capability of the residual heat removal by passive conduction cooldown from the core to the reactor cavity via an uninsulated vessel. The MHTGR uses a number of metallic components. A description of these components and their design requirements are presented in this paper.
Date: June 1, 1988
Creator: Shenoy, A.S. & Betts, W.S.
Partner: UNT Libraries Government Documents Department

Contain calculations of debris conditions adjacent to the BWR Mark I drywell shell during the later phases of a severe accident

Description: Best estimate CONTAIN calculations have recently been performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to predict the primary containment response during the later phases of an unmitigated low-pressure Short Term Station Blackout at the Peach Bottom Atomic Power Station. Debris pour conditions leaving the failed reactor vessel are taken from the results of best estimate BWRSAR analyses that are based upon an assumed metallic debris melting temperature of 2750/degree/F (1783 K) and an oxide debris melting temperature of 4350/degree/F (2672 K). Results of the CONTAIN analysis for the case without sprays indicate failure of the drywell seals due to the extremely hot atmospheric conditions extant in the drywell. The maximum calculated temperature of the debris adjacent to the drywell shell is less than the melting temperature of the shell, yet the sustained temperatures may be sufficient to induce primary containment pressure boundary failure by the mechanism of creep-rupture. It is also predicted that a significant portion of the reactor pedestal wall is ablated during the period of the calculation. Nevertheless, the calculated results are recognized to be influenced by large modeling uncertainties. Several deficiencies in the application of the CORCON module within the CONTAIN code to BWR severe accident sequences are identified and discussed. 5 refs., 9 figs., 4 tabs.,
Date: January 1, 1988
Creator: Hyman, C.R.
Partner: UNT Libraries Government Documents Department