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GT-MHR COMMERCIALIZATION STUDY FINAL CONTRACTUAL REPORT OF WORK PERFORMED FROM CONTRACT BEGINNING, JUNE 18,2001,TO CONTRACT END, JANUARY 31,2004

Description: OAK-B135 This is the final report of work performed by General Atomics on a Gas Turbine Modular Helium Reactor (GT-MHR) commercialization study under contract to the Department of Energy, Oakland Operations Office. The contract work scope covered a series of discrete tasks relating to commercialization of the GT-MHR. During the first year of performance, June 18, 2001--June 30, 2002, the contract covered six tasks, Tasks 1 through 6. Subsequently, four additional tasks were added, Tasks 7,8,10 and 11. With the exception of Task 1, each of the contract Tasks involved the development of one or more discrete deliverable products. Task 1 covered activities performed by General Atomics as part of a several year fuel irradiation testing activity being conducted in cooperation with the European Union. The irradiation testing will not be completed for three or more years into the future. Future work by General Atomics on this irradiation test activity will be covered by a new contract.
Date: February 1, 2004
Creator: SHENOY, AS
Partner: UNT Libraries Government Documents Department

Gas cooled fast reactor

Description: Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.
Date: June 1, 1972
Partner: UNT Libraries Government Documents Department

Coated Particle and Deep Burn Fuels Monthly Highlights December 2010

Description: During FY 2011 the CP & DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for November 2010, ORNL/TM-2010/323, was distributed to program participants on December 9, 2010. The final Quarterly for FY 2010, Deep Burn Program Quarterly Report for July - September 2010, ORNL/TM-2010/301, was announced to program participants and posted to the website on December 28, 2010. This report discusses the following: (1) Thermochemical Data and Model Development - (a) Thermochemical Modeling, (b) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Pebble Bed Design (INL), (c) Radiation Damage and Properties; (2) TRISO (tri-structural isotropic) Development - (a) TRU (transuranic elements) Kernel Development, (b) Coating Development; (3) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing (ORNL); (4) Fuel Performance and Analytical Analysis - Fuel Performance Modeling (ORNL).
Date: January 1, 2011
Creator: Snead, Lance Lewis; Bell, Gary L & Besmann, Theodore M
Partner: UNT Libraries Government Documents Department

Deep Burn Develpment of Transuranic Fuel for High-Temperature Helium-Cooled Reactors - July 2010

Description: The DB Program Quarterly Progress Report for April - June 2010, ORNL/TM/2010/140, was distributed to program participants on August 4. This report discusses the following: (1) TRU (transuranic elements) HTR (high temperature helium-cooled reactor) Fuel Modeling - (a) Thermochemical Modeling, (b) 5.3 Radiation Damage and Properties; (2) TRU HTR Fuel Qualification - (a) TRU Kernel Development, (b) Coating Development, (c) ZrC Properties and Handbook; and (3) HTR Fuel Recycle - (a) Recycle Processes, (b) Graphite Recycle, (c) Pyrochemical Reprocessing - METROX (metal recovery from oxide fuel) Process Development.
Date: August 1, 2010
Creator: Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D & Bell, Gary L
Partner: UNT Libraries Government Documents Department

Evaluation of the Gas Turbine Modular Helium Reactor

Description: Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs.
Date: February 1, 1994
Partner: UNT Libraries Government Documents Department

Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010

Description: The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).
Date: October 1, 2010
Creator: Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D & Bell, Gary L
Partner: UNT Libraries Government Documents Department

Optimized Flow Sheet for a Reference Commercial-Scale Nuclear-Driven High-Temperature Electrolysis Hydrogen Production Plant

Description: This report presents results from the development and optimization of a reference commercialscale high-temperature electrolysis (HTE) plant for hydrogen production. The reference plant design is driven by a high-temperature helium-cooled 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 consists of 4.176 × 10 6 cells with a per-cell active area of 225 cm2. A nominal cell area-specific resistance, ASR, value of 0.4 Ohm•cm2 with a current density of 0.25 A/cm2 was used, and isothermal boundary conditions were assumed. 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 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 overall system thermal-to-hydrogen production efficiency (based on the low heating value of the produced hydrogen) is 49.07% at a hydrogen production rate of 2.45 kg/s with the high-temperature helium-cooled reactor concept. The information presented in this report is intended to establish an optimized design for the reference nuclear-driven HTE hydrogen production plant so that parameters can be compared with other hydrogen production methods and power cycles to evaluate relative performance characteristics and plant economics.
Date: November 1, 2007
Creator: McKellar, M. G.; O'Brien, J. E.; Harvego, E. A. & Herring, J. S.
Partner: UNT Libraries Government Documents Department

Demonstration and System Analysis of High Temperature Steam Electrolysis for Large-Scale Hydrogen Production Using SOFCs

Description: At the Idaho National Engineering Laboratory, an integrated laboratory scale (ILS), 15 kW high-temperature electrolysis (HTE) facility has been developed under the U.S. Department of Energy Nuclear Hydrogen Initiative. Initial operation of this facility resulted in over 400 hours of operation with an average hydrogen production rate of approximately 0.9 Nm3/hr. The integrated laboratory scale facility is designed to address larger-scale issues such as thermal management (feed-stock heating, high-temperature gas handling), multiple-stack hot-zone design, multiple-stack electrical configurations, and other “integral” issues. Additionally, a reference process model of a commercial-scale high-temperature electrolysis plant for hydrogen production has been developed. The reference plant design is driven by a 600 megawatt thermal high-temperature helium-cooled reactor coupled to a direct Brayton power cycle. The electrolysis unit used to produce hydrogen consists of 4.01×106 cells with a per-cell active area of 225 cm2. A nominal cell area-specific resistance, ASR, value of 0.4 Ohm•cm2 with a current density of 0.25 A/cm2 was used, and isothermal boundary conditions were assumed. The overall system thermal-to-hydrogen production efficiency (based on the low heating value of the produced hydrogen) is 47.1% at a hydrogen production rate of 2.36 kg/s with the high-temperature helium-cooled reactor concept. This paper documents the initial operation of the ILS, with experimental details about heat-up, initial stack performance, as well as long-term operation and stack degradation. The paper will also present the optimized design for the reference nuclear-driven HTE hydrogen production plant which may be compared with other hydrogen production methods and power cycles to evaluate relative performance characteristics and plant economics.
Date: July 1, 2008
Creator: McKellar, Michael G.; O'Brien, James E.; Stoots, Carl M. & Herring, J. Stephen
Partner: UNT Libraries Government Documents Department

Sensitivity Studies of Advanced Reactors Coupled to High Temperature Electrolysis (HTE) Hydrogen Production Processes

Description: High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 °C to 950 °C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the steam or air sweep loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycle producing the highest efficiencies varied depending on the temperature range considered.
Date: April 1, 2007
Creator: Harvego, Edwin A.; McKellar, Michael G.; O'Brien, James E. & Herring, J. Stephen
Partner: UNT Libraries Government Documents Department

Conceptual Design of a Very High Temperature Pebble-Bed Reactor

Description: Efficient electricity and hydrogen production distinguish the Very High Temperature Reactor as the leading Generation IV advanced concept. This graphite-moderated, helium-cooled reactor achieves a requisite high outlet temperature while retaining the passive safety and proliferation resistance required of Generation IV designs. Furthermore, a recirculating pebble-bed VHTR can operate with minimal excess reactivity to yield improved fuel economy and superior resistance to ingress events. Using the PEBBED code developed at the INEEL, conceptual designs of 300 megawatt and 600 megawatt (thermal) Very High Temperature Pebble-Bed Reactors have been developed. The fuel requirements of these compare favorably to the South African PBMR. Passive safety is confirmed with the MELCOR accident analysis code.
Date: November 1, 2003
Creator: Gougar, Hans D.; Ougouag, A. M.; Moore, Richard M. & Terry, W. K.
Partner: UNT Libraries Government Documents Department

GCFR shielding design and supporting experimental programs

Description: The shielding for the conceptual design of the gas-cooled fast breeder reactor (GCFR) is described, and the component exposure design criteria which determine the shield design are presented. The experimental programs for validating the GCFR shielding design methods and data (which have been in existence since 1976) are also discussed.
Date: May 1, 1980
Creator: Perkins, R.G.; Hamilton, C.J. & Bartine, D.
Partner: UNT Libraries Government Documents Department

Integration of safety considerations into the design of GCFR safety systems

Description: Under DOE sponsorship, the GCFR Program is preparing a program to integrate reliability into the engineering and design of safety related systems, subsystems and components. This program is considered consistent with the NRC licensing position for CRBR which established a formal reliability objective for the prevention of core damage. The objective of the program is to provide assurance that reliability goals established for systems and subsystems are met consistent with the overall plant goals. Special consideration is given to components for which only a generic data base exists. Based on evaluations of past reliability test programs, it is concluded that full scale reliability test programs are not cost effective but that extended DV and S testing may be warranted in special circumstances. The major elements of the program, their relationship and benefits to the design of safety systems are discussed.
Date: October 1, 1979
Creator: Torri, A.; Kelley, A.P.; Emon, D.E. & Goetzmann, C.
Partner: UNT Libraries Government Documents Department

THE VALUE OF HELIUM-COOLED REACTOR TECHNOLOGIES OF NUCLEAR WASTE

Description: Helium-cooled reactor technologies offer significant advantages in accomplishing the waste transmutation process. They are ideally suited for use with thermal, epithermal, or fast neutron energy spectra. They can provide a relatively hard thermal neutron spectrum for transmutation of fissionable materials such as Pu-239 using ceramic-coated transmutation fuel particles, a graphite moderator, and a non-fertile burnable poison. These features (1) allow deep levels of transmutation with minimal or no intermediate reprocessing, (2) enhance passive decay heat removal via heat conduction and radiation, (3) allow operation at relatively high temperatures for a highly efficient generation of electricity, and (4) discharge the transmuted waste in a form that is highly resistant to corrosion for long times. They also offer the possibility for the use of epithermal neutrons that can interact with transmutable materials more effectively because of the large atomic cross sections in this energy domain. A fast spectrum may be useful for deep burnup of certain minor actinides. For this application, helium is essentially transparent to neutrons, does not degrade neutron energies, and offers the hardest possible neutron energy environment. In this paper, we report results from recent work on materials transmutation balances, safety, value to a geological repository, and economic considerations.
Date: March 1, 2001
Creator: RODRIGUEZ, C. & BAXTER, A.
Partner: UNT Libraries Government Documents Department

A feasibility study of reactor-based deep-burn concepts.

Description: A systematic assessment of the General Atomics (GA) proposed Deep-Burn concept based on the Modular Helium-Cooled Reactor design (DB-MHR) has been performed. Preliminary benchmarking of deterministic physics codes was done by comparing code results to those from MONTEBURNS (MCNP-ORIGEN) calculations. Detailed fuel cycle analyses were performed in order to provide an independent evaluation of the physics and transmutation performance of the one-pass and two-pass concepts. Key performance parameters such as transuranic consumption, reactor performance, and spent fuel characteristics were analyzed. This effort has been undertaken in close collaborations with the General Atomics design team and Brookhaven National Laboratory evaluation team. The study was performed primarily for a 600 MWt reference DB-MHR design having a power density of 4.7 MW/m{sup 3}. Based on parametric and sensitivity study, it was determined that the maximum burnup (TRU consumption) can be obtained using optimum values of 200 {micro}m and 20% for the fuel kernel diameter and fuel packing fraction, respectively. These values were retained for most of the one-pass and two-pass design calculations; variation to the packing fraction was necessary for the second stage of the two-pass concept. Using a four-batch fuel management scheme for the one-pass DB-MHR core, it was possible to obtain a TRU consumption of 58% and a cycle length of 286 EFPD. By increasing the core power to 800 MWt and the power density to 6.2 MW/m{sup 3}, it was possible to increase the TRU consumption to 60%, although the cycle length decreased by {approx}64 days. The higher TRU consumption (burnup) is due to the reduction of the in-core decay of fissile Pu-241 to Am-241 relative to fission, arising from the higher power density (specific power), which made the fuel more reactivity over time. It was also found that the TRU consumption can be improved by utilizing axial fuel shuffling ...
Date: September 16, 2005
Creator: Kim, T. K.; Taiwo, T. A.; Hill, R. N. & Yang, W. S.
Partner: UNT Libraries Government Documents Department

Refueling system for the gas-cooled fast breeder reactor

Description: Criteria specifically related to the handling of Gas-Cooled Fast Breeder Reactor (GCFR) fuel are briefly reviewed, and the most significant requirements with which the refueling system must comply are discussed. Each component of the refueling system is identified, and a functional description of the fuel handling machine is presented. An illustrated operating sequence describing the various functions involved in a typical refueling cycle is presented. The design status of components and subsystems selected for conceptual development is reviewed, and anticipated refueling time frames are given.
Date: May 1, 1980
Creator: Hawke, B.C.
Partner: UNT Libraries Government Documents Department

GCFR radial blanket and shield experiment: objectives, preanalysis, and specifications

Description: An integral experiment has been designed for the verification of radiation transport methods and nuclear data used for the design of the radial shield for the proposed 300 MW(e) gas-cooled fast breeder reactor (GCFR). The scope of the experiment was chosen to include a thorium oxide radial blanket mockup as well as several shield configurations in order to reduce the uncertainties in the calculated source terms for the radial shield, and to reduce the uncertainties in the calculated radiation damage to the prestressed concrete reactor vessel (PCRV). Additionally, the measurements are intended to bound the uncertainties in calculated gamma-ray heating rates within the blanket and shield. Although designed specifically for the GCFR, the experiment will provide generic data regarding deep penetration in ThO/sub 2/ and common shield materials, which should also benefit LMFBR designers.
Date: September 1, 1979
Creator: Ingersoll, D.T.; Bartine, D.E. & Muckenthaler, F.J.
Partner: UNT Libraries Government Documents Department

Overview of the GCFR core engineering development program

Description: The gas-cooled fast breeder reactor (GCFR) core development plan has been described as an incremental extension of the LMFBR program. The basis for this contention is reviewed along with some additional gas-cooled reactor precedence. Ideally, in the development of a reactor concept, it would be desirable to demonstrate the perforance of a full-sized fuel assembly under the anticipated irradiation and coolant conditions. Practically, however, this is seldom possible, particularly in fast reactors where fluence-to-burnup ratios increase with reactor power. Because of this limitation, the GCFR development program is based on qualifying the assemblies for the low distortions that occur during the swelling incubation period (approx. 50 MWd/kg). Extended burnup will be demonstrated in the operation of the GCFR demonstration plant core. The overall core development program and the related assumptions are discussed and the design goals stated. A multistage development approach is described and illustratd with an abbreviated program network. Major areas of the development program are briefly described.
Date: July 1, 1979
Creator: Snyder, H.J. Jr.
Partner: UNT Libraries Government Documents Department

Helium-cooled high temperature reactors

Description: Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg.
Date: January 1, 1985
Creator: Trauger, D.B.
Partner: UNT Libraries Government Documents Department

Analysis of gas-cooled fast reactor shield designs

Description: In its shielding program for the Gas-Cooled Fast Reactor (GCFR) as conceived by General Atomic, Oak Ridge National Laboratory has developed an advanced shielding analysis system that incorporates the latest analysis techniques for converging to a shield design compatible with other design parameters such as cooling and structural requirements or material compatibility. Basically the system consists in applying the various techniques in a logical sequence to a given design, thereby generating a large body of data to serve as an information base for subsequent redesign. As an illustration, this system is applied to successive typical models for the GCFR, resulting in a reduction in the thickness of the radial shield and redesign of the lower shield region. In principle, the design-analysis-redesign iterations would continue until they converge upon an acceptable configuration.
Date: January 1, 1978
Creator: Bartine, D.E. & Williams, L.R.
Partner: UNT Libraries Government Documents Department

Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

Description: The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.
Date: September 1, 1978
Partner: UNT Libraries Government Documents Department

FSV experience in support of the GT-MHR reactor physics, fuel performance, and graphite

Description: The Fort St. Vrain (FSV) power plant was the most recent operating graphite-moderated, helium-cooled nuclear power plant in the United States. Many similarities exist between the FSV design and the current design of the GT-MHR. Both designs use graphite as the basic building blocks of the core, as structural material, in the reflectors, and as a neutron moderator. Both designs use hexagonal fuel elements containing cylindrical fuel rods with coated fuel particles. Helium is the coolant and the power densities vary by less than 5%. Since material and geometric properties of the GT-MHR core am very similar to the FSV core, it is logical to draw upon the FSV experience in support of the GT-MHR design. In the Physics area, testing at FSV during the first three cycles of operation has confirmed that the calculational models used for the core design were very successful in predicting the core nuclear performance from initial cold criticality through power operation and refueling. There was excellent agreement between predicted and measured initial core criticality and control rod positions during startup. Measured axial flux distributions were within 5% of the predicted value at the peak. The isothermal temperature coefficient at zero power was in agreement within 3%, and even the calculated temperature defect over the whole operating range for cycle 3 was within 8% of the measured defect. In the Fuel Performance area, fuel particle coating performance, and fission gas release predictions and an overall plateout analysis were performed for decommissioning purposes. A comparison between predicted and measured fission gas release histories of Kr-85m and Xe-138 and a similar comparison with specific circulator plateout data indicated good agreement between prediction and measured data. Only I-131 plateout data was overpredicted, while Cs-137 data was underpredicted.
Date: November 1, 1994
Creator: Baxter, A. M.; McEachern, D.; Hanson, D. L. & Vollman, R. E.
Partner: UNT Libraries Government Documents Department

EVALUATION OF ZERO-POWER, ELEVATED-TEMPERATURE MEASUREMENTS AT JAPAN’S HIGH TEMPERATURE ENGINEERING TEST REACTOR

Description: The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The experimental benchmark performed and currently evaluated in this report pertains to the data available for two zero-power, warm-critical measurements with the fully-loaded HTTR core. Six isothermal temperature coefficients for the fully-loaded core from approximately 340 to 740 K have also been evaluated. These experiments were performed as part of the power-up tests (References 1 and 2). Evaluation of the start-up core physics tests specific to the fully-loaded core (HTTR-GCR-RESR-001) and annular start-up core loadings (HTTR-GCR-RESR-002) have been previously evaluated.
Date: March 1, 2011
Creator: Bess, John D.; Fujimoto, Nozomu; Sterbentz, James W.; Snoj, Luka & Zukeran, Atsushi
Partner: UNT Libraries Government Documents Department

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

Description: The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis ...
Date: October 1, 2011
Creator: Pope, Michael A.
Partner: UNT Libraries Government Documents Department