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Experimental plan for irradiation experiment HRB-21

Description: Irradiation experiment HRB-21 is the first in a series of test capsules that are designed to provide a fuel-performance data base to be used for the validation of modular high-temperature gas-cooled reactor (MHTGR) coated-particle fuel performance models under MHTGR normal operating conditions and specific licensing basis events. Capsule HRB-21 will contain an advanced TRISO-P UCO/ThO{sub 2} - coated-particle fuel system with demonstrated low defective-particle fraction ({le}5 {times} 10{sup {minus}5}) and a heavy metal-contamination fraction ({le}1 {times} 10{sup {minus}5}) that meets MHTGR quality specifications. The coated particles and fuel compacts were fabricated in laboratory-scale facilities using MHTGR reference procedures at General Atomics (GA). Nearly 150,000 fissile and fertile particles will be irradiated in capsule HRB-21 at a mean volumetric fuel temperature of 975{degree}C and will achieve a peak fissile burnup of 26% fissions per initial metal atom (FIMA) while accumulating a fast neutron fluence of about 4.5 {times} 10{sup 25} neutrons/m{sup 2}. This experiment is a cooperative effort between the US Department of Energy (DOE) and the Japan Atomic Energy Research Institute (JAERI). The participants are the Oak Ridge National Laboratory (ORNL), GA, and the Tokai Research Establishment. Capsule HRB-21 will contain the US MHTGR fuel specimens, and a companion capsule, HRB-22, will contain the JAERI fuel. The irradiation will take place in the removable beryllium reflector facility of the High Flux Isotope Reactor (HFIR) at ORNL. The performance of the fuel during irradiation will be closely monitored through on-line fission gas release measurements. Detailed postirradiation examination and conduction cooldown simulation testing will be performed on the irradiated fuel compacts from both the HRB-21 and HRB-22 capsules. 5 refs., 9 figs., 6 tabs.
Date: April 1, 1989
Creator: Goodin, D. T.; Kania, M. J. & Patton, B. W.
Partner: UNT Libraries Government Documents Department

Core fluctuations test. Revision 1

Description: Fluctuations were first encountered in the Fort St. Vrain reactor early in cycle 1 operation, during the initial rise from 40% to 70% power. Subsequent in-core tests and operation throughout cycles 1 and 2 demonstrated that fluctuations were repeatable, occurring at core pressure drops of between 2.5 psi and 4.0 psi, and that in each instance their characteristics were very similar. Subsequently, tests and analysis were done to understand the core fluctuation phenomenon. These efforts also lead to a design fix which stopped these fluctuations in the FSV reactor core. This fix required that keys be used in addition to the keys in the core support floor which already existed. This report outlines a test plan to validate that core fluctuations will not occur in the MHTGR core. 2 refs., 12 figs., 3 tabs.
Date: June 1, 1987
Creator: Betts, W.S.
Partner: UNT Libraries Government Documents Department

Probabilistic risk assessment of the modular HTGR plant. Revision 1

Description: A preliminary probabilistic risk assessment (PRA) has been performed for the modular HTGR (MHTGR). This PRA is preliminary in the context that although it updates the PRA issued earlier to include a wider spectrum of events for Licensing Basis Events (LBE) selection, the final version will not be issued until later. The primary function of the assessment was to assure compliance with the NRC interim safety goals imposed by the top-level regulatory criteria, and utility/user requirements regarding public evacuation or sheltering. In addition, the assessment provides a basis for designer feedback regarding reliability allocations and barrier retention requirements as well as providing a basis for the selection of licensing basis events (LBEs) and the safety classification of structures, systems, and components. The assessment demonstrates that both the NRC interim safety goals and utility/user imposed sheltering/evacuation requirements are satisfied. Moreover, it is not anticipated that design changes introduced will jeopardize compliance with the interim safety goals or utility/user requirements. 61 refs., 48 figs., 24 tabs.
Date: June 1, 1986
Creator: Everline, C.J.; Bellis, E.A. & Vasquez, J.
Partner: UNT Libraries Government Documents Department

Vessel support subsystem design description. Revision 1

Description: The Vessel Support Subsystem is one of three subsystems comprising the Vessel System of the Modular High Temperature Gas-Cooled Reactor 4 x 350 MW(t) Plant. The design of this subsystem has been developed by means of the Integrated Approach. This document establishes the functions and system design requirements of the Vessel Support Subsystem from the Functional Analysis, and includes institutional requirements from the Overall Plant Design Specification and the Vessel System Design Description. A description of the subsystem design which satisfies these requirements is presented. Lower-tier requirements at the subsystem level are next defined for the component design. This document also includes information on aspects of subsystem construction, operation, maintenance, and decommissioning.
Date: July 1, 1987
Creator: Perry, R.A. & Mehta, D.D.
Partner: UNT Libraries Government Documents Department

350 MW(t) design fuel cycle selection. Revision 1

Description: This document discusses the results of this evaluation and a recommendation to retain the graded fuel cycle in which one-half of the fuel elements are exchanged at each refueling. This recommendation is based on the better performance of the graded cycle relative to the evaluation criteria of both economics and control margin. A choice to retain the graded cycle and a power density of 5.9 MW/m{sup 3} for the upcoming conceptual design phase was deemed prudent for the following reasons: the graded cycle has significantly better economics, and essentially the same expected availability factor as the batch design, when both are evaluated against the same requirements, including water ingress; and the reduction in maximum fuel pin power peaking in the batch design compared to the graded cycle is only a few percent and gas hot streaks are not improved by changing to a batch cycle. The preliminary 2-D power distribution studies for both designs showed that maximum fuel pin power peaking, particularly near the inner reflector, was high for both designs and nearly the same in magnitude. 10 figs., 9 tabs.
Date: January 1986
Creator: Lane, R. K.; Lefler, W. & Shirley, G.
Partner: UNT Libraries Government Documents Department

Design data needs modular high-temperature gas-cooled reactor. Revision 2

Description: The Design Data Needs (DDNs) provide summary statements for program management, of the designer`s need for experimental data to confirm or validate assumptions made in the design. These assumptions were developed using the Integrated Approach and are tabulated in the Functional Analysis Report. These assumptions were also necessary in the analyses or trade studies (A/TS) to develop selections of hardware design or design requirements. Each DDN includes statements providing traceability to the function and the associated assumption that requires the need.
Date: March 1, 1987
Partner: UNT Libraries Government Documents Department

Defect fractions for fissile and fertile TRISO-coated fuel

Description: High quality TRISCO-coated UCO and ThO{sub 2} particles with reference MHTGR dimensions were produced in a coating campaign in August and September 1986 for irradiation tests. The heavy metal contamination and the defect levels were below the limits established for the MHTGR fuel. Over 9 kg of uranium in UCO and 30 kg of thorium in ThO{sub 2} were TRISCO-coated in 4 fissile and 3 fertile batches in the 240mm Development Coater. These coated fuel particles will be used to produce fuel rods for testing in the irradiation validation tests to be conducted in capsules HRB-19, -20 and -21 on the DOE Fuel and Fission Product Technology Program. 3 refs., 6 figs., 6 tabs.
Date: September 1, 1986
Creator: Adams, C.C.
Partner: UNT Libraries Government Documents Department

Helium Storage and Transfer Subsystem design description. Revision

Description: The Helium Storage and Transfer Subsystem (HSTS) consists of two parts. The first consists of nine (9) high pressure storage tanks containing helium at 15.6 MPa (2250 psig). These tanks provide makeup and purge helium at a rate of 1216 kg per y (2680 lb/y) to the various helium users, including circulator bearing seals, analysis packages, and cooling system surge tanks. The second, larger part of the system, provides for the low pressure storage of 6078 kg (13,400 lb) of primary coolant helium in 180 storage tanks at 7.0 MPa (1000 psig). The system serves all four (4) reactor modules. The low pressure storage part of the system receives helium from the discharge of Helium Purification Subsystem (HPS) and is activated during depressurization and pumpup operations only. It is not required to operate continuously. Storage capacity is provided for primary helium coolant from two reactor modules. However, since depressurization and pumpup operations are performed for only one reactor module at a time, two 50% capacity low pressure transfer compressors are provided having a total transfer capacity of 340 am{sup 3}/h (200 acfm) which is sufficient to service one module. High pressure helium is supplied continuously to all the four reactor modules simultaneously from the high pressure storage tanks. These tanks are replaced periodically with fresh tanks.
Date: July 1, 1987
Partner: UNT Libraries Government Documents Department

Security monitoring subsystem design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant

Description: Security Monitoring acquires and processes sensor data for use by security personnel in the performance of their function. Security Monitoring is designed and implemented as a part of an overall security plan which is classified as Safeguards Information under 10CFR73.21.
Date: June 1, 1986
Partner: UNT Libraries Government Documents Department

Overall plant design specification Modular High Temperature Gas-cooled Reactor. Revision 9

Description: Revision 9 of the ``Overall Plant Design Specification Modular High Temperature Gas-Cooled Reactor,`` DOE-HTGR-86004 (OPDS) has been completed and is hereby distributed for use by the HTGR Program team members. This document, Revision 9 of the ``Overall Plant Design Specification`` (OPDS) reflects those changes in the MHTGR design requirements and configuration resulting form approved Design Change Proposals DCP BNI-003 and DCP BNI-004, involving the Nuclear Island Cooling and Spent Fuel Cooling Systems respectively.
Date: May 1, 1990
Partner: UNT Libraries Government Documents Department

Evaluation of need for integral fuel oxidation tests

Description: This document establishes the need for an integral fuel oxidation test which can give confidence to the predictions made by the OXIDE computer code for fuel and core damage during water ingress events in the Modular High Temperature Gas Cooled Reactor (MHTGR). This testing will provide clear engineering evidence to demonstrate that the core of the MHTGR can survive a moisture ingress incident with minimum investment risk and without danger to the reactor personnel or to the public. In particular, these tests will determine the degree of particle debonding and compact stack densification as a function of the fractional compact matrix burnoff. Also included in the document is a description of the proposed tests, and, a test matrix of the planned experiments. 3 refs., 1 fig., 1 tab.
Date: February 1, 1987
Creator: Montgomery, F.C.
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

Site fuel handling subsystem design description. Revision

Description: The Site Fuel Handling Subsystem (SFHS) consists of equipment and facilities located in the reactor Service Building which are used to handle hexagonal graphite fuel and reflector blocks. This equipment interfaces closely with the core refueling equipment. The SFHS uses some of the equipment in the Core Refueling System to transfer fuel elements between the spent fuel storage facility (part of Core Refueling Subsystem, HFD-43413) and the fuel sealing and inspection facility (FSIF).
Date: July 1, 1987
Partner: UNT Libraries Government Documents Department

Containment vs confinement trade study, small HTGR plant PCRV [prestressed concrete reactor vessel] concept

Description: This trade study has been conducted to evaluate the differences between four different HTGR nuclear power plants. All of the plants use a prestressed concrete reactor vessel (PCRV) to house the core and steam generation equipment. The reactor uses LEU U/Th fuel in prismatic carbon blocks. All plant concepts meet the utility/user requirements established for small HTGR plants. All plants will be evaluated with regard to their ability to produce safe, economical power to satisfy Goals 1, 2, and 3 of the HTGR program and by meeting the MUST criteria established in the concept evaluation plan. Capital costs for each plant were evaluated on a differential cost basis. These costs were developed according to the ``NUS`` code of accounts as defined in the Cost Estimating and Control Procedure, HP-20901. Accounts that were identical in scope for all four plants were not used for the comparison. Table 1-1 is a list of capital cost accounts that were evaluated for each plant.
Date: March 1, 1985
Partner: UNT Libraries Government Documents Department

Report on chemical impurities in lots of 2020 graphite

Description: Chemical and physical studies on agglomerated and dispersed impurities found in three lots of 2020 graphite received from the Stackpole Company are described. Each lot consisted of three billets, one shipment of off-the-shelf 2020 and two of ``nuclear quality`` graphite for which efforts had been taken to decrease impurity levels. An earlier report described the results of chemical analyses of the three lots of graphite and the gross distribution of the impurities in the billets. The present report describes the chemical and physical studies on the detailed distribution and chemical identity of the agglomerates that led to the identification of their source. Agglomerated impurities were found in all lots of 2020. The particles possessed an average size range of about 0.3 mm. In addition to these particulates, other inorganic ash forming components were found to be rather more uniformly distributed internally throughout the graphite. 4 refs., 4 figs., 3 tabs.
Date: November 1, 1985
Creator: Strehlow, R.A.
Partner: UNT Libraries Government Documents Department

Modular high temperature gas-cooled reactor plant design duty cycle. Revision 3

Description: This document defines the Plant Design Duty Cycle (PCDC) for the Modular High Temperature Gas-cooled Reactor (MHTGR). The duty cycle is a set of events and their design number of occurrences over the life of the plant for which the MHTGR plant shall be designed to ensure that the plant meets all the top-level requirements. The duty cycle is representative of the types of events to be expected in multiple reactor module-turbine plant configurations of the MHTGR. A synopsis of each PDDC event is presented to provide an overview of the plant response and consequence. 8 refs., 1 fig., 4 tabs.
Date: December 31, 1989
Creator: Chan, T.
Partner: UNT Libraries Government Documents Department

Operating procedure for SiC defect detection: Data support document

Description: The feasibility of the Hg Intrusion QC method for measuring SiC coating defects for the MHTGR was conducted as a potential improvement for the Burn/Leach (B/L) QC method currently used. The purpose for evaluating the Hg Intrusion QC method as an alternative method was to determine if B/L QC method underestimated SiC coating defects. Some evidence in work conducted earlier, indicated that TRISO-coated fuel particles with low SiC coating defects measured by the B/L QC method showed higher releases of metallic fission products. These data indicated that the SiC coating defect fractions were higher than the B/L measured data indicated. Sample sizes used in the current study were too small to conclusively demonstrate that the B/L QC method under estimate SiC coating defects. However, observations made during this study indicated a need for an additional QC method to the B/L QC method to measure SiC coating defects for the higher quality MHTGR fuels. The B/L QC method is the best method for measuring SiC coating defects with missing SiC layers or broken SiC coatings (gross SiC defects). However, SiC coating defects with microcracks and other SiC defects not detected by the B/L method may contribute to the release of metallic fission products in-service. For these type of SiC coating defects, the Hg Intrusion QC method investigated in this study is feasible, but particle sample size should be increased to a much larger sample size (100,000 particles per test) for the MHTGR. 7 refs., 5 figs., 9 tabs.
Date: September 29, 1989
Creator: Adams, C. C. & Partain, K. E.
Partner: UNT Libraries Government Documents Department

Hot Service Facility subsystem design description. Revision

Description: The Hot Service Facility Subsystem, which is also referred to as the Reactor Equipment Service Facility (RESF), is located in an environmentally controlled shielded vault and provides inspection, maintenance, care, and repair of reactor service equipment and tools. The shielded vault is located in the Reactor Service Building.
Date: July 1, 1987
Partner: UNT Libraries Government Documents Department

Maintenance building structural design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant

Description: The Maintenance Building is a grade-founded, two-story, steel-framed structure, located adjacent to the Turbine Building in the Energy Conversion Area. It has a reinforced concrete foundation and slab on grade, and insulated sheet metal exterior walls and roof decking.
Date: June 1, 1986
Partner: UNT Libraries Government Documents Department

HVAC [Heating, Ventilation and Air Conditioning] subsystem design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant

Description: The HVAC system is a subsystem within the Mechanical Services Group (MSG). The HVAC system for the 4 x 350 MW(t) Modular HTGR Plant presently consists of ten, nonsafety-related subsystems located in the Nuclear Island (NI) and Energy Conversion Area (ECA) of the plant.
Date: June 1, 1986
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

Solid radioactive waste subsystem design description

Description: The Solid Radioactive Waste Subsystem provides reliable processing of collected solid waste to meet the requirements of 10CFR20 and 10CFR61. The methods utilized are cement solidification for sludges, resins, liquids, and noncompactible waste, and compaction for dry compressible waste. The drums of processed waste will be stored until transported off-site for disposal at a licensed burial site.
Date: June 1, 1986
Partner: UNT Libraries Government Documents Department

Nuclear Island Engineering MHTGR [Modular High-Temperature Gas-cooled Reactor] preliminary and final designs. Technical progress report, December 12, 1988--September 30, 1989

Description: This report summarizes the Department of Energy (DOE)-funded work performed by General Atomics (GA) under the Nuclear Island Engineering (NIE)-Modular High-Temperature Gas-cooled Reactor (MHTGR) Preliminary and Final Designs Contract DE-AC03-89SF17885 for the period December 12, 1988 through September 30, 1989. This reporting period is the first (partial) fiscal year of the 5-year contract performance period. The objective of DOE`s MHTGR program is to advance the design from the conceptual design phase into preliminary design and then on to final design in support of the Nuclear Regulatory Commission`s (NRC`s) design review and approval of the MHTGR Design Team, is focused on the Nuclear Island portion of the technology and design, primarily in the areas of the reactor and internals, fuel characteristics and fuel fabrication, helium services systems, reactor protection, shutdown cooling, circulator design, and refueling system. Maintenance and implementation of the functional methodology, plant-level analysis, support for probabilistic risk assessment, quality assurance, operations, and reliability/availability assessments are included in GA`s scope of work.
Date: December 1, 1989
Partner: UNT Libraries Government Documents Department