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Cracked pellet gap conductance model: comparison of FRAP-S calculations with measured fuel centerline temperatures

Description: Fuel pellets crack extensively upon irradiation due both to thermal stresses induced by power changes and at high burnup, to accumulation of gaseous fission products at grain boundaries. Therefore, the distance between the fuel and cladding will be circumferentially nonuniform; varying between that calculated for intact operating fuel pellets and essentially zero (fuel segments in contact with the cladding wall). A model for calculation of temperatures in cracked pellets is proposed wherein the effective fuel to cladding gap conductance is calculated by taking a zero pressure contact conductance in series with an annular gap conductance. Comparisons of predicted and measured fuel centerline temperatures at beginning of life and at extended burnup are presented in support of the model. 13 references. (auth)
Date: March 1, 1975
Creator: MacDonald, P.E. & Broughton, J.M.
Partner: UNT Libraries Government Documents Department

In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

Description: The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made.
Date: January 1, 1980
Creator: Gunnerson, F.S. & MacDonald, P.E.
Partner: UNT Libraries Government Documents Department

Blowdown mass flow measurements during the Power Burst Facility LOC-11C test

Description: An interpretation and evaluation of the two-phase coolant mass flow measurements obtained during Test LOC-11C performed in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL) are presented. Although a density gradient existed within the pipe between 1 and 6 s, the homogeneous flow model used to calculate the coolant mass flow from the measured mixture density, momentum flux, and volumetric flow was found to be generally satisfactory. A cross-sectional average density was determined by fitting a linear density gradient through the upper and lower chordal densities obtained from a three-beam gamma densitometer and then combining the result with the middle beam density. The integrated measured coolant mass flow was subsequently found to be within 5% if the initial mass inventory of the PBF loss-of-coolant accident (LOCA) system. The posttest calculations using the RELAP4/MOD6 computer code to determine coolant mass flow for Test LOC-11C also agreed well with the measured data.
Date: January 1, 1979
Creator: Broughton, J.M. & MacDonald, P.E.
Partner: UNT Libraries Government Documents Department

Safety and licensing issues that are being addressed by the Power Burst Facility test programs. [PWR; BWR]

Description: This paper presents an overview of the results of the experimental program being conducted in the Power Burst Facility and the relationship of these results to certain safety and licensing issues. The safety issues that were addressed by the Power-Cooling-Mismatch, Reactivity Initiated Accident, and Loss of Coolant Accident tests, which comprised the original test program in the Power Burst Facility, are discussed. The resolution of these safety issues based on the results of the thirty-six tests performed to date, is presented. The future resolution of safety issues identified in the new Power Burst Facility test program which consists of tests which simulate BWR and PWR operational transients, anticipated transients without scram, and severe fuel damage accidents, is described.
Date: January 1, 1980
Creator: McCardell, R.K. & MacDonald, P.E.
Partner: UNT Libraries Government Documents Department

PBF experimental program

Description: The fuels behavior research in PBF is directed towards providing a detailed understanding of the response of nuclear fuel assemblies to off-normal and hypothetical accident conditions. Single fuel rods and clusters of highly instrumented fuel rods are installed within a central test space of the PBF core for testing. The core can be operated in various modes to provide test conditions typical of accidents and off-normal conditions such as: (1) Power-Cooling-Mismatch (PCM) accidents; (2) Loss-of-Coolant Accident (LOCA); (3) Reactivity-Initiated-Accident (RIA); (4) Operational Transients With and Without Scram (OPTRAN); (5) Small Break LOCAs (SBL). This document provides an overview of each of the PBF test series with emphasis on what has been learned to date and why the remaining tests are being conducted. (DLC)
Date: January 1, 1979
Creator: MacDonald, P.E. & Zeile, H.J.
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Research and Development Program Plan

Description: The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission Demonstrate safe and economical nuclearassisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented in Section 4. ...
Date: January 1, 2005
Creator: MacDonald, P. E.
Partner: UNT Libraries Government Documents Department

Analysis of fuel behavior during reactivity initiated accidents

Description: An analysis of fuel rod behavior during reactivity initiated accidents is presented. The calculational approach is described, predictions are discussed and compared to data from the SPERT power excursion tests, and the sensitivity of maximum cladding temperatures to initial reactor conditions, enthalpy insertions, and other experiment design parameters is investigated. 10 references. (auth)
Date: March 1, 1975
Creator: Thompson, L.B.; Tolman, E.L. & MacDonald, P.E.
Partner: UNT Libraries Government Documents Department

Evaluation of the thermal-hydraulic response and fuel rod thermal and mechanical deformation behavior during the power burst facility test LOC-3. [PWR]

Description: An evaluation of the results from the LOC-3 nuclear blowdown test conducted in the Power Burst Facility is presented. The test objective was to examine fuel and cladding behavior during a postulated cold leg break accident in a pressurized water reactor (PWR). Separate effects of rod internal pressure and the degree of irradiation were investigated in the four-rod test. Extensive cladding deformation (ballooning) and failure occurred during blowdown. The deformation of the low and high pressure rods was similar; however, the previously irradiated test rod deformed to a greater extent than a similar fresh rod exposed to identical system conditions.
Date: January 1, 1980
Creator: Yackle, T.R.; MacDonald, P.E. & Broughton, J.M.
Partner: UNT Libraries Government Documents Department

Comparison of measured and calculated LWR fuel behavior during a hypothetical reactivity initiated accident

Description: Comparisons of measured and calculated LWR fuel rod responses during a reactivity initiated accident test are presented. The results indicate that the computer code, FRAP-T5, adequately calculates the fuel rod behavior up to the time at which the gap closes and provides a good thermal solution up to the time of gross fuel and cladding relocation. Three areas have been identified for further model development: (a) development of a fuel swelling model for near molten fuel temperatures; (b) incorporation of a deformable fuel pellet model that is applicable for the closed gap, high strain rate regime; and (c) expansion of the failure stress data base to include heat-up rates of up to 5000 K/s.
Date: January 1, 1980
Creator: Fukuda, S.K.; MacDonald, P.E. & Garner, R.W.
Partner: UNT Libraries Government Documents Department

LWR fuel rod behavior observed during postulated accident conditions: a comparison of FRAP-T calculated and PBF experimental results

Description: Light water reactor (LWR) fuel rod behavior during transient experiments conducted in the Power Burst Facility is reviewed. The experiments examined simulated hypothetical reactivity initiated accidents (RIA) and power-cooling-mismatch (PCM) events. Fuel rod behavior calculated by the Fuel Rod Analysis Program-Transient (FRAP-T) is compared with the test data. Important physical phenomena observed during the tests and not presently incorporated into the FRAP-T code are: (a) fuel swelling in the radial direction due to fission gas effects, (b) UO/sub 2/-zircaloy chemical interaction, and (c) loss of UO/sub 2/ grain boundary strength and fuel powdering. Additional models needed in FRAP-T to reflect the fuel behavior observed during the two types of transients are cladding thickness variation during an RIA, molten fuel movement and possible cladding-molten fuel thermal interaction during a PCM event, and in the case of breached rods, the effects of hydrogen pickup on cladding embrittlement.
Date: January 1, 1980
Creator: Charyulu, M.K.; Hobbins, R.R. & MacDonald, P.E.
Partner: UNT Libraries Government Documents Department

The Next Generation Nuclear Plant (NGNP) Project

Description: The Next Generation Nuclear Power (NGNP) Project will demonstrate emissions-free nuclearassisted electricity and hydrogen production by 2015. The NGNP reactor will be a helium-cooled, graphite moderated, thermal neutron spectrum reactor with a design goal outlet temperature of 1000 C or higher. The reactor thermal power and core configuration will be designed to assure passive decay heat removal without fuel damage during hypothetical accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. This paper provides a description of the project to build the NGNP at the Idaho National Engineering and Environmental Laboratory (INEEL). The NGNP Project includes an overall reactor design activity and four major supporting activities: materials selection and qualification, NRC licensing and regulatory support, fuel development and qualification, and the hydrogen production plant. Each of these activities is discussed in the paper. All the reactor design and construction activities will be managed under the DOE’s project management system as outlined in DOE Order 413.3. The key elements of the overall project management system discussed in this paper include the client and project management organization relationship, critical decisions (CDs), acquisition strategy, and the project logic and timeline. The major activities associated with the materials program include development of a plan for managing the selection and qualification of all component materials required for the NGNP; identification of specific materials alternatives for each system component; evaluation of the needed testing, code work, and analysis required to qualify each identified material; preliminary selection of component materials; irradiation of needed sample materials; physical, mechanical, and chemical testing of unirradiated and irradiated materials; and documentation of final materials selections. The NGNP will be licensed by the NRC under 10 CFR 50 or 10 CFR 52, for the purpose of demonstrating the suitability of high-temperature gas-cooled reactors for commercial electric ...
Date: November 1, 2003
Creator: Southworth, F. H. & MacDonald, P. E.
Partner: UNT Libraries Government Documents Department

USNRC-OECD Halden Project fuel behavior test program: experiment data report for test assemblies IFA-226 and IFA 239

Description: The experimental data which were obtained from the IFA-226 and IFA-239 test assemblies during operation in the Halden Boiling Water Reactor are reported. Included are cladding elongation, fuel centerline temperature, internal gas pressure, and power history data from IFA-226 which were obtained from November 1971 through April 1974, and cladding elongation, diametral profile, and power history data from IFA-239 covering the period from March 1973 through April 1974. The data, presented in the form of composite graphs, have been analyzed only to the extent necessary to assure that they are reasonable and correct. A description of these mixed oxide fuel test assemblies and their instrumentation is presented. Test pin fabrication history, instrument calibration data, assembly power calibration methods, and the neutron detector data reduction technique are included as appendices. (auth)
Date: December 1, 1975
Creator: Laats, E.T.; MacDonald, P.E. & Quapp, W.J.
Partner: UNT Libraries Government Documents Department

Advanced nondestructive examination technologies for measuring fatigue damage in nuclear power plant components

Description: This paper presents recent results from an ongoing project at the Idaho National Engineering Laboratory (INEL) to develop advanced nondestructive methods to characterize the aging degradation of nuclear power plant pressure boundary components. One of the advanced methods, positron annihilation, is being developed for in situ characterization of fatigue damage in nuclear power plant piping and other components. This technique can detect and correlate the microstructural changes that are precursors of fatigue cracking in austenitic stainless steel components. In fact, the initial INEL test results show that the method can detect fatigue damage in stainless steel ranging from a few percent of the fatigue life up to 40 percent.
Date: December 1, 1995
Creator: MacDonald, P.E.; Shah, V.N. & Akers, D.W.
Partner: UNT Libraries Government Documents Department

Characteristics of a Mixed Thorium-Uranium Dioxide High-Burnup Fuel

Description: Future nuclear fuels must satisfy three sets of requirements: longer times between refueling; concerns for weapons proliferation; and development of a spent fuel form more suitable for direct geologic disposal. This project has investigated a fuel consisting of mixed thorium and uranium dioxide to satisfy these requirements. Results using the SCALE 4.3 code system indicated that the mixed Th-U fuel could be burned to 72 MWD/kg or 100 MWD/kg using 25% of 35% UO2 respectively. The uranium remained below 20% total fissile fraction throughout the cycle, making it unusable for weapons. Total plutonium production per MWD was a factor of 4.5 less in the Th-U fuel than in the conventional fuel; Pu-239 production per MWD was a factor of 6.5 less; and the plutonium produced was high in Pu-238, leading to a decay heat 5 times greater than that from plutonium derived from conventional fuel and 40 times greater than weapons grade plutonium. High decay heat would require active cooling of any crude weapon, lest the components surrounding the plutonium be melted. Spontaneous neutron production for plutonium from Th-U fuel was 2.3 times greater than that from conventional fuel and 15 times greater than that from weapons grade plutonium. High spontaneous neutron production drastically limits the probable yield of a crude weapon. Because ThO2 is the highest oxide of thorium, while UO2 can be oxidized further to U3O8, ThO2-UO2 fuel may be a superior wasteform if the spent fuel is ever to be exposed to oxygenated water. Even if the cost of fabricating the mixed Th-U fuel is $100/kg greater, the cost of the Th-U fuel is 13% to 15% less than that of the fuels using uranium only.
Date: June 1, 1999
Creator: Herring, J. S. & MacDonald, P. E.
Partner: UNT Libraries Government Documents Department

TMI-2 core examination

Description: The examination of the damaged core at the Three Mile Island Unit 2 (TMI-2) reactor is structured to address the following safety issues: fission product release, transport, and deposition; core coolability; containment integrity; and recriticality during severe accidents; as well as zircaloy cladding ballooning and oxidation during so-called design basis accidents. The numbers of TMI-2 components or samples to be examined, the priority of each examination, the safety issue addressed by each examination, the principal examination techniques to be employed, and the data to be obtained and the principal uses of the data are discussed in this paper.
Date: January 1, 1983
Creator: Hobbins, R.R.; MacDonald, P.E. & Owen, D.E.
Partner: UNT Libraries Government Documents Department

Thermal-hydraulics of the PFB/LOFT lead rod loss-of-coolant experiments. [PWR]

Description: Results of the four PBF/LOFT Lead Rod sequential blowdown tests conducted in the Power Burst Facility (PBF) are presented. The primary objective of the test series was to evaluate the extent of mechanical deformation that would be expected to occur to low pressure (0.1 MPa), light water reactor design fuel rods subjected to a series of nuclear blowdown tests, and to determine if subjecting deformed fuel rods to subsequent testing would result in rod failure. The extent of mechanical deformation (buckling, collapse, or waisting of the cladding) was evaluated by comparison of cladding temperature versus system pressure response with out-of-pile experimental data, and by posttest visual examinations and cladding diametral measurements. Tests LLR-3, LLR-5, LLR-4, and LLR-4A were performed at system conditions of 595/sup 0/K coolant inlet temperature, 15.5 MPa system pressure, and 41, 46, 57 and 56 kW/m test rod peak linear powers, respectively, at initiation of blowdown. Cladding temperatures during the tests ranged from 870 to 1260/sup 0/K.
Date: January 1, 1980
Creator: Varacalle, D.J. Jr.; Garner, R.W.; MacDonald, P.E. & Cox, W.R.
Partner: UNT Libraries Government Documents Department

Fuel rod mechanical deformation during the PBF/LOFT lead rod loss-of-coolant experiments

Description: Results of four PBF/LOFT Lead Rod (LLR) sequential blowdown tests conducted in the Power Burst Facility (PBF) are presented. Each test employed four separately shrouded fuel rods. The primary objective of the test series was to evaluate the extent of mechanical deformation that would be expected to occur to low pressure (0.1 MPa), light water reactor design fuel rods when subjected to a series of double ended cold leg break loss-of-coolant accident (LOCA) tests, and to determine whether subjecting these deformed fuel rods to subsequent testing would result in rod failure. The extent of mechanical deformation (buckling, collapse, or waisting of the cladding) was evaluated by comparison of cladding temperature and system pressure measurements with out-of-pile experimental data, and by posttest visual examinations and cladding diametral measurements.
Date: January 1, 1980
Creator: Varacalle, D.J. Jr.; MacDonald, P.E.; Shiozawa, S. & Driskell, W.E.
Partner: UNT Libraries Government Documents Department

Light water reactor fuel response during reactivity initiated accident experiments

Description: Experimental results from six recent Power Burst Facility (PBF) reactivity initiated accident (RIA) tests are compared with data from previous Special Power Excursion Reactor Test (SPERT), and Japanese Nuclear Safety Research Reactor (NSRR) tests. The RIA fuel behavior experimental program recently started in the PBF is being conducted with coolant conditions typical of hot-startup conditions in a commercial boiling water reactor. The SPERT and NSRR test programs investigated the behavior of single or small clusters of light water reactor (LWR) type fuel rods under approximate room temperature and atmospheric pressure conditions in capsules containing stagnant water. As observed in the SPERT and NSRR tests, energy deposition, and consequent enthalpy increase in the PBF test fuel, appears to be the single most important variable. However, the consequences of failure at boiling water hot-startup system conditions appear to be more severe than previously observed in either the stagnant capsule SPERT or NSRR tests. Metallographic examination of both previously unirradiated and irradiated PBF fuel rod cross sections revealed extensive variation in cladding wall thicknesses (involving considerable plastic flow) and fuel shattering along grain boundaries in both restructured and unrestructured fuel regions. Oxidation of the cladding resulted in fracture at the location of cladding thinning and disintegration of the rods during quench. In addition,swelling of the gaseous and potentially volatile fission products in previously irradiated fuel resulted in volume increases of up to 180% and blockage of the coolant channels within the flow shrouds surrounding the fuel rods.
Date: January 1, 1979
Creator: MacDonald, P.E.; McCardell, R.K.; Martinson, Z.R. & Seiffert, S.L.
Partner: UNT Libraries Government Documents Department

Technical basis for the proposed high efficiency nuclear fuel program

Description: Greenhouse gas emissions from fossil fired electricity generating stations will dramatically increase over the next 20 years. Nuclear energy is the only fully developed technology able to supply large amounts of electricity without generation of greenhouse gases. However, the problem of noncompetitive economics and public concerns about radioactive waste disposal, safety, and nuclear weapons proliferation may prevent the reemergence of nuclear power as a preferred option for new electric energy generation in the US. This paper discusses a new research program to help address these issues, by developing fuel designs capable of burnup values in excess of 60 MWD/kgU. The objectives of the program are to: Improve the reliability and robustness of light water reactor fuel, thereby improving safety margins; significantly increase the energy generated by each fuel loading, thereby achieving longer operating cycles, higher capacity factors, and lower cost electric power; significantly reduce the volume of spent nuclear fuel discharged for disposal by allowing more energy to be extracted from each fuel element prior to discharge; and develop fuel that is much more proliferation resistant.
Date: July 1, 1998
Creator: MacDonald, P.E.; Herring, J.S.; Crawford, D.C. & Neimark, L.E.
Partner: UNT Libraries Government Documents Department

Steam generator tube failures

Description: A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service.
Date: April 1, 1996
Creator: MacDonald, P.E.; Shah, V.N.; Ward, L.W. & Ellison, P.G.
Partner: UNT Libraries Government Documents Department

Evidence of aging effects on certain safety-related components

Description: In response to interest shown by the Nuclear Energy Agency (NEA), Principal Working Group I (PWG- 1) of the Committee on the Safety of Nuclear Installations (CSNI) conducted a generic study on the effects of aging of active components in nuclear power plants. (This focus on active components is consistent with PWG-l`s mandate; passive components are primarily within the mandate of PWG-3.) Representatives from France, Sweden, Finland, Japan, the United States, and the United Kingdom participated in the study by submitting reports documenting aging studies performed in their countries. This report consists of summaries of those reports, along with a comparison of the various statistical analysis methods used in the studies. The studies indicate that with some exceptions, active components generally do not present a significant aging problem in nuclear power plants. Design criteria and effective preventative maintenance programs, including timely replacement of components, are effective in mitigating potential aging problems. However, aging studies (such as qualitative and statistical analyses of failure modes and maintenance data) are an important part of efforts to identify and solve potential aging problems. Solving these problems typically includes such strategies as replacing suspect components with improved components, and implementing improved maintenance programs.
Date: January 1, 1996
Creator: Magleby, H.L.; Atwood, C.L.; MacDonald, P.E.; Edson, J.L. & Bramwell, D.L.
Partner: UNT Libraries Government Documents Department

Light water reactor fuel response during reactivity initiated accident experiments

Description: Experimental results from six recent Power Burst Facility (PBF) reactivity initiated accident (RIA) tests are compared with data from previous SPERT, TREAT and NSRR programs. The RIA fuel behavior experimental program recently started in the PBF is being conducted with coolant conditions typical of hot-startup conditions in a commercial boiling water reactor. The SPERT, TREAT and NSRR test programs investigated the behavior of single or small clusters of light water reactor (LWR) type fuel rods under approximate room temperature and atmospheric pressure conditions in capsules containing stagnant water. As observed in the previous tests, energy depositions, and consequent enthalpy increase, in the PBF test fuel appears to be the single most important variable. However, the consequences of failure at boiling water hot-startup system conditions appear to be more severe than previously observed in either the stagnant SPERT or NSRR tests. Metallographic examination of both previously unirradiated and irradiated PBF fuel rod cross sections revealed extensive variation in cladding wall thickness involving considerable plastic flow and fuel shattering along grain boundaries in both restructured and unrestructured fuel regions. In addition, swelling of the gaseous and potentially volatile fission products in previously irradiated fuel resulted in volume increases of up to 180% and blockage of the flow shrouds surrounding the fuel rods.
Date: January 1, 1979
Creator: MacDonald, P.E.; McCardell, R.K.; Martinson, Z.R.; Hobbins, R.R.; Seiffert, S.L. & Cook, B.A.
Partner: UNT Libraries Government Documents Department

Assessment of light water reactor fuel damage during a reactivity initiated accident

Description: This paper presents an assessment of LWR fuel damage during a reactivity initiated accident and comments on the adequacy of the present USNRC design requirements. Results from early SPERT tests are reviewed and compared with results from recent computer simulations and PBF tests. A progression of fuel rod and cladding damage events is presented. High strain rate deformation of relatively cool irradiated cladding early in the transient may result in fracture at a radial average peak fuel enthalpy of approximately 140 cal/g UO/sub 2/. Volume expansion of previously irradiated fuel upon melting may cause deformation and rupture of the cladding, and coolant channel blockage at higher peak enthalpies.
Date: January 1, 1980
Creator: MacDonald, P.E.; Seiffert, S.L.; Martinson, Z.R.; McCardell, R.K.; Owen, D.E. & Fukuda, S.K.
Partner: UNT Libraries Government Documents Department

Very High Temperature Reactor (VHTR) Survey of Materials Research and Development Needs to Support Early Deployment

Description: The VHTR reference concept is a helium-cooled, graphite moderated, thermal neutron spectrum reactor with an outlet temperature of 1000 C or higher. It is expected that the VHTR will be purchased in the future as either an electricity producing plant with a direct cycle gas turbine or a hydrogen producing (or other process heat application) plant. The process heat version of the VHTR will require that an intermediate heat exchanger (IHX) and primary gas circulator be located in an adjoining power conversion vessel. A third VHTR mission - actinide burning - can be accomplished with either the hydrogen-production or gas turbine designs. The first ''demonstration'' VHTR will produce both electricity and hydrogen using the IHX to transfer the heat to either a hydrogen production plant or the gas turbine. The plant size, reactor thermal power, and core configuration will be designed to assure passive decay heat removal without fuel damage during accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. The purpose of this report is to identify the materials research and development needs for the VHTR. To do this, we focused on the plant design described in Section 2, which is similar to the GT-MHR plant design (850 C core outlet temperature). For system or component designs that present significant material challenges (or far greater expense) there may be some viable design alternatives or options that can reduce development needs or allow use of available (cheaper) materials. Nevertheless, we were not able to assess those alternatives in the time allotted for this report and, to move forward with this material research and development assessment, the authors of this report felt that it was necessary to use a GT-MHR type design as the baseline design.
Date: January 1, 2003
Creator: Shaber, Eric; Baccaglini, G.; Ball, S.; Burchell, T.; Corwin, B.; Fewell, T. et al.
Partner: UNT Libraries Government Documents Department