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Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project

Description: This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.
Date: January 26, 2007
Creator: Culp, A. B.
Partner: UNT Libraries Government Documents Department

An Engineering Test Reactor

Description: A relatively inexpensive reactor for the specific purpose of testing a sub-critical portion of another reactor under conditions that would exist during actual operation is discussed. It is concluded that an engineering tool for reactor development work that bridges the present gap between exponential and criticality experiments and the actual full scale operating reactor is feasible. An example of such a test reactor which would not entail development effort to ut into operation is depicted.
Date: March 16, 1951
Creator: Fahrner, T.; Stoker, R.L. & Thomson, A.S.
Partner: UNT Libraries Government Documents Department

Estimation of Critical Flow Velocity for Collapse of Gas Test Loop Booster Fuel Assembly

Description: This paper presents calculations performed to determine the critical flow velocity for plate collapse due to static instability for the Gas Test Loop booster fuel assembly. Long, slender plates arranged in a parallel configuration can experience static divergence and collapse at sufficiently high coolant flow rates. Such collapse was exhibited by the Oak Ridge High Flux Reactor in the 1940s and the Engineering Test Reactor at the Idaho National Laboratory in the 1950s. Theoretical formulas outlined by Miller, based upon wide-beam theory and Bernoulli’s equation, were used for the analysis. Calculations based upon Miller’s theory show that the actual coolant flow velocity is only 6% of the predicted critical flow velocity. Since there is a considerable margin between the theoretically predicted plate collapse velocity and the design velocity, the phenomena of plate collapse due to static instability is unlikely.
Date: July 1, 2006
Creator: Guillen & Russell, Mark J.
Partner: UNT Libraries Government Documents Department

NGNP Fuel Qualification White Paper

Description: The Japanese high temperature gas reactor program is centered on the High Temperature Engineering Test Reactor (HTTR), which has a thermal power of 30 MW and 950°C maximum coolant outlet temperature. The HTTR achieved criticality in November 1998 and has undergone a series of rise-to-power tests [Fujikawa 2004]. In December 2001, an outlet temperature of 850°C was achieved and in April 2004 a temperature of 950°C was achieved. As of July 2004, the reactor had operated for 224 effective full power days (EFPD). The planned core life cycle is 660 EFPD [Verfondern 2000]. It is planned to couple a high temperature process heat application to the HTTR through its intermediate heat exchanger in the future.
Date: July 1, 2010
Creator: Petti, David A.
Partner: UNT Libraries Government Documents Department

Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

Description: The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.
Date: October 1, 1994
Partner: UNT Libraries Government Documents Department

Criticality Benchmark Analysis of the HTTR Annular Startup Core Configurations

Description: One of the high priority benchmarking activities for corroborating the Next Generation Nuclear Plant (NGNP) Project and Very High Temperature Reactor (VHTR) Program is evaluation of Japan's existing High Temperature Engineering Test Reactor (HTTR). The HTTR is a 30 MWt engineering test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. A large amount of critical reactor physics data is available for validation efforts of High Temperature Gas-cooled Reactors (HTGRs). Previous international reactor physics benchmarking activities provided a collation of mixed results that inaccurately predicted actual experimental performance.1 Reevaluations were performed by the Japanese to reduce the discrepancy between actual and computationally-determined critical configurations.2-3 Current efforts at the Idaho National Laboratory (INL) involve development of reactor physics benchmark models in conjunction with the International Reactor Physics Experiment Evaluation Project (IRPhEP) for use with verification and validation methods in the VHTR Program. Annular cores demonstrate inherent safety characteristics that are of interest in developing future HTGRs.
Date: November 1, 2009
Creator: Bess, John D.
Partner: UNT Libraries Government Documents Department

Engineering Evaluation/Cost Analysis for Decommissioning of the Engineering Test Reactor Complex

Description: Preparation of this Engineering Evaluation/Cost Analysis is consistent with the joint U.S. Department of Energy and U.S. Environmental Protection Agency Policy on Decommissioning of Department of Energy Facilities Under the Comprehensive Environmental Response, Compensation, and Liability Act, which establishes the Comprehensive Environmental Response, Compensation, and Liability Act non-time-critical removal action (NTCRA) process as an approach for decommissioning.
Date: October 1, 2006
Creator: Culp, A. B.
Partner: UNT Libraries Government Documents Department

Irradiation Facilities at the Advanced Test Reactor

Description: The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC – formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world’s data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities1. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens.
Date: December 1, 2005
Creator: Grover, S. Blaine
Partner: UNT Libraries Government Documents Department

Reducing Electrical Power Use with a Performance Based Incentive

Description: This Departmental Energy Management Program (DEMP) funded Model Program Study developed out of a potential DOE-ID Performance Based Incentive for the Idaho National Engineering and Environmental Laboratory (INEEL), lasting from October 2001 through May 2002, which stressed reductions in electrical usage. An analysis of demand usage obtained from monthly INEEL Power Management electric reports revealed reductions in demand from a majority of the site areas. The purpose of this Model Program study was to determine the methods and activities that were used at these site areas to achieve the reductions in demand and to develop these demand reduction methods and activities into a Model Program that could be shared throughout the INEEL and DOE complex-wide for additional demand savings. INEEL Energy Management personnel interviewed contacts from the eight areas which had achieved a consistent reduction in demand during the study period, namely, Idaho Nuclear Technology and Engineering Center (INTEC), Test Area North (TAN), Power Burst Facility (PBF), Test Reactor Area (TRA) including Advanced Test Reactor ATR), Engineering Test Reactor (ETR), and Materials Test Reactor (MTR) areas, Central Facilities Area (CFA), Specific Manufacturing Capability (SMC), Radioactive Waste Management Complex (RWMC), and Argonne National Laboratory-West (ANLW). The information that resulted from the interviews indicated that more than direct demand and energy reduction actions were responsible for the recorded reductions in demand. INEEL Energy Management identified five categories of actions or conditions that contributed to the demand reduction. These categories are Decontamination and Decommissioning (D&D), employee actions, improvements, inactivation for maintenance, and processes. The following information details the findings from the study.
Date: July 1, 2004
Creator: Nell, M. Kathleen
Partner: UNT Libraries Government Documents Department

WDC-1-1 instrumented irradiation of boron carbide in a spectrum-hardened ETR flux

Description: Boron carbide pellets were irradiated in a spectrum-hardened ETR flux at temperatures of 1220 to 1620 deg F up to maximum specimen-averaged exposures of 38 x 10/sup 20/ captures/cm/sup 3/. Material variables included pellet density (75 to 99% T. D.) and stoichiometry (B: C = 3.8 to 4.1). Gas release and irradiation temperature were continuously monitored during the irradiation. Results are reported for boron carbide structural degradation, helium release, and interaction with Type 316 stainless steel. (auth)
Date: April 1, 1973
Creator: Pitner, A.L.
Partner: UNT Libraries Government Documents Department

Benchmark Development in Support of Generation-IV Reactor Validation (IRPhEP 2010 Handbook)

Description: The March 2010 edition of the International Reactor Physics Experiment Evaluation Project (IRPhEP) Handbook includes additional benchmark data that can be implemented in the validation of data and methods for Generation IV (GEN-IV) reactor designs. Evaluations supporting sodium-cooled fast reactor (SFR) efforts include the initial isothermal tests of the Fast Flux Test Facility (FFTF) at the Hanford Site, the Zero Power Physics Reactor (ZPPR) 10B and 10C experiments at the Idaho National Laboratory (INL), and the burn-up reactivity coefficient of Japan’s JOYO reactor. An assessment of Russia’s BFS-61 assemblies at the Institute of Physics and Power Engineering (IPPE) provides additional information for lead-cooled fast reactor (LFR) systems. Benchmarks in support of the very high temperature reactor (VHTR) project include evaluations of the HTR-PROTEUS experiments performed at the Paul Scherrer Institut (PSI) in Switzerland and the start-up core physics tests of Japan’s High Temperature Engineering Test Reactor. The critical configuration of the Power Burst Facility (PBF) at the INL which used ternary ceramic fuel, U(18)O2-CaO-ZrO2, is of interest for fuel cycle research and development (FCR&D) and has some similarities to “inert-matrix” fuels that are of interest in GEN-IV advanced reactor design. Two additional evaluations were revised to include additional evaluated experimental data, in support of light water reactor (LWR) and heavy water reactor (HWR) research; these include reactor physics experiments at Brazil’s IPEN/MB-01 Research Reactor Facility and the French High Flux Reactor (RHF), respectively. The IRPhEP Handbook now includes data from 45 experimental series (representing 24 reactor facilities) and represents contributions from 15 countries. These experimental measurements represent large investments of infrastructure, experience, and cost that have been evaluated and preserved as benchmarks for the validation of methods and collection of data in support of current and future reactor design and development.
Date: June 1, 2010
Creator: Bess, John D. & Briggs, J. Blair
Partner: UNT Libraries Government Documents Department

Processing data for coextruded fuel elements in ETR, KER and MTR Loops

Description: The following data represents processing conditions used in fabricating prototypic (except for supports) natural and enriched NPR element assemblies and K-type I&E{sup 2} for the following production tests: ETR, 6{times}6 Loop; GEH-10, Nos. 52, 53, 54 and 55; KER Loops; PT-377; MTR Loop; GEH-4, Nos. 68, 69 and 70. The purpose in documenting the data is to provide a permanent record of processing conditions and dimensions which may be referred to for post irradiation analysis and possible future process work. Post irradiation results will be issued by the Fuels Development Operation, Hanford Laboratories Operation, and the test loop operating conditions will be issued by Process and Reactor Development Operation, Irradiation Processing Department, as outlined in the Production Test Procedure.
Date: November 8, 1961
Creator: Robinson, R. K.
Partner: UNT Libraries Government Documents Department

In-reactor rupture testing of Zircaloy-2 clad seven-rod cluster fuel elements

Description: Three tests have been run in the ETR in high temperature, high pressure, recirculating water. In one test, previously unirradiated fuel elements were used and in the other two the fuel was irradiated to 2400 MWD/T at HAPO prior to insertion in the ETR. Failure was initiated by shearing off a projection on the surface of one rod of a fuel element, thus opening a 25-mil hole through the cladding. The projection was sheared off by a hydraulically operated chisel controlled from outside the reactor. The first test was operated seven hours after the defect was opened with no failure. Failure is defined as having occurred when sufficient uranium oxide has formed to split open the cladding and release large amounts of fission products into the loop water. The second test was operated for fourteen hours after the defect was opened with again no failure. The third test was operated for only 33 minutes after the defect cap was sheared off before fission product activity in the loop water caused the test to be terminated.
Date: April 15, 1960
Creator: Call, R. L. & Kaulitz, D. C.
Partner: UNT Libraries Government Documents Department

In-reactor rupture testing of Zircaloy-2 clad seven-rod cluster fuel elements: Interim report

Description: The high pressure loop installed in the 3X3 reflector position of the ETR and the associated instrumentation to detect and study failure mechanisms handled the rupture tests without difficulty. Failure of the elements was initiated by shearing off a projection on the fuel elements. The first test of the series used previously unirradiated seven-rod clusters. After the projection was sheared off the fuel elements were operated for seven hours with no failure. Failure is defined as having occurred when sufficient uranium oxide has formed to split open the cladding and release large amounts of fission products into the loop water. The second and third tests used fuel which had been irradiated to 2400 MWD/T at Hanford prior to insertion into the ETR. The second test was operated for 14 hours after the projection was sheared off--again with no failure. The third test was operated for only 33 minutes after the projection was sheared off before fission product activity in the loop water caused the test to be terminated.
Date: May 3, 1960
Creator: Call, R. L.; Green, J. W. & Kaulitz, D. C.
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

Preliminary Benchmark Evaluation of Japan’s High Temperature Engineering Test Reactor

Description: A benchmark model of the initial fully-loaded start-up core critical of Japan’s High Temperature Engineering Test Reactor (HTTR) was developed to provide data in support of ongoing validation efforts of the Very High Temperature Reactor Program using publicly available resources. The HTTR is a 30 MWt test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. The benchmark was modeled using MCNP5 with various neutron cross-section libraries. An uncertainty evaluation was performed by perturbing the benchmark model and comparing the resultant eigenvalues. The calculated eigenvalues are approximately 2-3% greater than expected with an uncertainty of ±0.70%. The primary sources of uncertainty are the impurities in the core and reflector graphite. The release of additional HTTR data could effectively reduce the benchmark model uncertainties and bias. Sensitivity of the results to the graphite impurity content might imply that further evaluation of the graphite content could significantly improve calculated results. Proper characterization of graphite for future Next Generation Nuclear Power reactor designs will improve computational modeling capabilities. Current benchmarking activities include evaluation of the annular HTTR cores and assessment of the remaining start-up core physics experiments, including reactivity effects, reactivity coefficient, and reaction-rate distribution measurements. Long term benchmarking goals might include analyses of the hot zero-power critical, rise-to-power tests, and other irradiation, safety, and technical evaluations performed with the HTTR.
Date: May 1, 2009
Creator: Bess, John Darrell
Partner: UNT Libraries Government Documents Department

BENCHMARK EVALUATION OF THE START-UP CORE REACTOR PHYSICS MEASUREMENTS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR

Description: The benchmark evaluation of the start-up core reactor physics measurements performed with Japan’s High Temperature Engineering Test Reactor, in support of the Next Generation Nuclear Plant Project and Very High Temperature Reactor Program activities at the Idaho National Laboratory, has been completed. The evaluation was performed using MCNP5 with ENDF/B-VII.0 nuclear data libraries and according to guidelines provided for inclusion in the International Reactor Physics Experiment Evaluation Project Handbook. Results provided include updated evaluation of the initial six critical core configurations (five annular and one fully-loaded). The calculated keff eigenvalues agree within 1s of the benchmark values. Reactor physics measurements that were evaluated include reactivity effects measurements such as excess reactivity during the core loading process and shutdown margins for the fully-loaded core, four isothermal temperature reactivity coefficient measurements for the fully-loaded core, and axial reaction rate measurements in the instrumentation columns of three core configurations. The calculated values agree well with the benchmark experiment measurements. Fully subcritical and warm critical configurations of the fully-loaded core were also assessed. The calculated keff eigenvalues for these two configurations also agree within 1s of the benchmark values. The reactor physics measurement data can be used in the validation and design development of future High Temperature Gas-cooled Reactor systems.
Date: May 1, 2010
Creator: Bess, John Darrell
Partner: UNT Libraries Government Documents Department

Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)

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 Japanese government approved construction of the HTTR in the 1989 fiscal year budget; construction began at the Oarai Research and Development Center in March 1991 and was completed May 1996. Fuel loading began July 1, 1998, from the core periphery. The first criticality was attained with an annular core on November 10, 1998 at 14:18, followed by a series of start-up core physics tests until a fully-loaded core was developed on December 16, 1998. Criticality tests were carried out into January 1999. The first full power operation with an average core outlet temperature of 850ºC was completed on December 7, 2001, and operational licensing of the HTTR was approved on March 6, 2002. The HTTR attained high temperature operation at 950 ºC in April 19, 2004. After a series of safety demonstration tests, it will be used as the heat source in a hydrogen production system by 2015. Hot zero-power critical, rise-to-power, irradiation, and safety demonstration testing , have also been performed with the HTTR, representing additional means for computational validation efforts. Power tests were performed in steps from 0 ...
Date: March 1, 2010
Creator: Bess, John D.; Fujimoto, Nozomu; Sterbentz, James W.; Snoj, Luka & Zukeran, Atsushi
Partner: UNT Libraries Government Documents Department