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Chemical effects head-loss research in support of generic safety issue 191.

Description: This summary report describes studies conducted at Argonne National Laboratory on the potential for chemical effects on head loss across sump screens. Three different buffering solutions were used for these tests: trisodium phosphate (TSP), sodium hydroxide, and sodium tetraborate. These pH control agents used following a LOCA at a nuclear power plant show various degrees of interaction with the insulating materials Cal-Sil and NUKON. Results for Cal-Sil dissolution tests in TSP solutions, settling rate tests of calcium phosphate precipitates, and benchmark tests in chemically inactive environments are also presented. The dissolution tests were intended to identify important environmental variables governing both calcium dissolution and subsequent calcium phosphate formation over a range of simulated sump pool conditions. The results from the dissolution testing were used to inform both the head loss and settling test series. The objective of the head loss tests was to assess the head loss produced by debris beds created by Cal-Sil, fibrous debris, and calcium phosphate precipitates. The effects of both the relative arrival time of the precipitates and insulation debris and the calcium phosphate formation process were specifically evaluated. The debris loadings, test loop flow rates, and test temperature were chosen to be reasonably representative of those expected in plants with updated sump screen configurations, although the approach velocity of 0.1 ft/s used for most of the tests is 3-10 times that expected in plants with large screens . Other variables were selected with the intent to reasonably bound the head loss variability due to arrival time and calcium phosphate formation uncertainty. Settling tests were conducted to measure the settling rates of calcium phosphate precipitates (formed by adding dissolved Ca to boric acid and TSP solutions) in water columns having no bulk directional flow. For PWRs where NaOH and sodium tetraborate are used to control sump ...
Date: October 31, 2006
Creator: Park, J. H.; Kasza, K.; Fisher, B.; Oras, J.; Natesan, K.; Shack, W. J. et al.
Partner: UNT Libraries Government Documents Department

ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

Description: We describe the next generation general purpose Evaluated Nuclear Data File, ENDF/B-VII.0, of recommended nuclear data for advanced nuclear science and technology applications. The library, released by the U.S. Cross Section Evaluation Working Group (CSEWG) in December 2006, contains data primarily for reactions with incident neutrons, protons, and photons on almost 400 isotopes. The new evaluations are based on both experimental data and nuclear reaction theory predictions. The principal advances over the previous ENDF/B-VI library are the following: (1) New cross sections for U, Pu, Th, Np and Am actinide isotopes, with improved performance in integral validation criticality and neutron transmission benchmark tests; (2) More precise standard cross sections for neutron reactions on H, {sup 6}Li, {sup 10}B, Au and for {sup 235,238}U fission, developed by a collaboration with the IAEA and the OECD/NEA Working Party on Evaluation Cooperation (WPEC); (3) Improved thermal neutron scattering; (4) An extensive set of neutron cross sections on fission products developed through a WPEC collaboration; (5) A large suite of photonuclear reactions; (6) Extension of many neutron- and proton-induced reactions up to an energy of 150 MeV; (7) Many new light nucleus neutron and proton reactions; (8) Post-fission beta-delayed photon decay spectra; (9) New radioactive decay data; and (10) New methods developed to provide uncertainties and covariances, together with covariance evaluations for some sample cases. The paper provides an overview of this library, consisting of 14 sublibraries in the same, ENDF-6 format, as the earlier ENDF/B-VI library. We describe each of the 14 sublibraries, focusing on neutron reactions. Extensive validation, using radiation transport codes to simulate measured critical assemblies, show major improvements: (a) The long-standing underprediction of low enriched U thermal assemblies is removed; (b) The {sup 238}U, {sup 208}Pb, and {sup 9}Be reflector biases in fast systems are largely removed; (c) ENDF/B-VI.8 good ...
Date: October 2, 2006
Creator: Chadwick, M B; Oblozinsky, P; Herman, M; Greene, N M; McKnight, R D; Smith, D L et al.
Partner: UNT Libraries Government Documents Department

Tabulation of Fundamental Assembly Heat and Radiation Source Files

Description: The purpose of this calculation is to tabulate a set of computer files for use as input to the WPLOAD thermal loading software. These files contain details regarding heat and radiation from pressurized water reactor (PWR) assemblies and boiling water reactor (BWR) assemblies. The scope of this calculation is limited to rearranging and reducing the existing file information into a more streamlined set of tables for use as input to WPLOAD. The electronic source term files used as input to this calculation were generated from the output files of the SAS2H/ORIGIN-S sequence of the SCALE Version 4.3 modular code system, as documented in References 2.1.1 and 2.1.2, and are included in Attachment II.
Date: October 25, 2006
Creator: deBues, T. & Ryman, J.C.
Partner: UNT Libraries Government Documents Department

Advanced Test Reactor Capabilities and Future Irradiation Plans

Description: The Advanced Test Reactor (ATR), located at the Idaho National Laboratory (INL), is one of the most versatile operating research reactors in the Untied States. The ATR has a long history of supporting reactor fuel and material research for the US government and other test sponsors. The INL is owned by the US Department of Energy (DOE) and currently operated by Battelle Energy Alliance (BEA). The ATR is the third generation of test reactors built at the Test Reactor Area, now named the Reactor Technology Complex (RTC), whose mission is to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The current experiments in the ATR are for a variety of customers--US DOE, foreign governments and private researchers, and commercial companies that need neutrons. The ATR has several unique features that enable the reactor to perform diverse simultaneous tests for multiple test sponsors. The ATR has been operating since 1967, and is expected to continue operating for several more decades. The remainder of this paper discusses the ATR design features, testing options, previous experiment programs, future plans for the ATR capabilities and experiments, and some introduction to the INL and DOE's expectations for nuclear research in the future.
Date: October 1, 2006
Creator: Marshall, Frances M.
Partner: UNT Libraries Government Documents Department

REACTOR PRESSURE VESSEL TEMPERATURE ANALYSIS OF CANDIDATE VERY HIGH TEMPERATURE REACTOR DESIGNS

Description: Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code. Because PEBBED-THERMIX has not been extensively validated, confirmatory calculations were also performed with RELAP5-3D for the pebble-bed design. During normal operation, the predicted axial profiles in reactor vessel temperature were similar with both codes and the predicted maximum values were within 2 °C. The trends of the calculated vessel temperatures were similar during the depressurized conduction cooldown accident. The maximum value predicted with RELAP5-3D during the depressurized conduction cooldown accident was about 40 °C higher than that predicted with PEBBED. This agreement is considered reasonable based on the expected uncertainty in either calculation. The differences between the PEBBED and RELAP5-3D calculations were not large enough to affect conclusions concerning comparisons between calculated and allowed maximum temperatures during normal operation and the depressurized conduction cooldown accident.
Date: October 1, 2006
Creator: Gougar, Hans D.; Davis, Cliff B.; Hayner, George & Weaver, Kevan
Partner: UNT Libraries Government Documents Department

Safety Assurance for ATR Irradiations

Description: The Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL) is the world’s premiere test reactor for performing high fluence, large volume, irradiation test programs. The ATR has many capabilities and a wide variety of tests are performed in this truly one of a kind reactor, including isotope production, simple self-contained static capsule experiments, instrumented/controlled experiments, and loop testing under pressurized water conditions. Along with the five pressurized water loops, ATR may also have gas (temperature controlled) lead experiments, fuel boosted fast flux experiments, and static sealed capsules all in the core at the same time. In addition, any or all of these tests may contain fuel or moderating materials that can affect reactivity levels in the ATR core. Therefore the safety analyses required to ensure safe operation of each experiment as well as the reactor itself are complex. Each test has to be evaluated against stringent reactor control safety criteria, as well as the effects it could have on adjacent tests and the reactor as well as the consequences of those effects. The safety analyses of each experiment are summarized in a document entitled the Experiment Safety Assurance Package (ESAP). The ESAP references and employs the results of the reactor physics, thermal, hydraulic, stress, seismic, vibration, and all other analyses necessary to ensure the experiment can be irradiated safely in the ATR. The requirements for reactivity worth, chemistry compatibilities, pressure limitations, material issues, etc. are all specified in the Technical Safety Requirements and the Upgraded Final Safety Analysis Report (UFSAR) for the ATR. This paper discusses the ESAP process, types of analyses, types of safety requirements and the approvals necessary to ensure an experiment can be safely irradiated in the ATR.
Date: October 1, 2006
Creator: Grover, S. Blaine
Partner: UNT Libraries Government Documents Department

The Modular Helium Reactor for Hydrogen Production

Description: For electricity and hydrogen production, an advanced reactor technology receiving considerable international interest is a modular, passively-safe version of the high-temperature, gas-cooled reactor (HTGR), known in the U.S. as the Modular Helium Reactor (MHR), which operates at a power level of 600 MW(t). For hydrogen production, the concept is referred to as the H2-MHR. Two concepts that make direct use of the MHR high-temperature process heat are being investigated in order to improve the efficiency and economics of hydrogen production. The first concept involves coupling the MHR to the Sulfur-Iodine (SI) thermochemical water splitting process and is referred to as the SI-Based H2-MHR. The second concept involves coupling the MHR to high-temperature electrolysis (HTE) and is referred to as the HTE-Based H2-MHR.
Date: October 1, 2006
Creator: Harvego, E.; Richards, M.; Shenoy, A.; Schultz, K.; Brown, L. & Fukuie, M.
Partner: UNT Libraries Government Documents Department

Testing of Gas Reactor Fuel and Materials in the Advanced Test Reactor

Description: The recent growth in interest for high temperature gas reactors has resulted in an increased need for materials and fuel testing for this type of reactor. The Advanced Test Reactor (ATR), located at the US Department of Energy’s Idaho National Laboratory, has long been involved in testing gas reactor fuel and materials, and has facilities and capabilities to provide the right environment for gas reactor irradiation experiments. These capabilities include both passive sealed capsule experiments, and instrumented/actively controlled experiments. The instrumented/actively controlled experiments typically contain thermocouples and control the irradiation temperature, but on-line measurements and controls for pressure and gas environment have also been performed in past irradiations. The ATR has an existing automated gas temperature control system that can maintain temperature in an irradiation experiment within very tight bounds, and has developed an on-line fission product monitoring system that is especially well suited for testing gas reactor particle fuel. The ATR’s control system, which consists primarily of vertical cylinders used to rotate neutron poisons/reflectors toward or away from the reactor core, provides a constant vertical flux profile over the duration of each operating cycle. This constant chopped cosine shaped axial flux profile, with a relatively flat peak at the vertical centre of the core, is more desirable for experiments than a constantly moving axial flux peak resulting from a control system of axially positioned control components which are vertically withdrawn from the core.
Date: October 1, 2006
Creator: Grover, S. Blaine
Partner: UNT Libraries Government Documents Department

PEBBLES: A COMPUTER CODE FOR MODELING PACKING, FLOW AND RECIRCULATIONOF PEBBLES IN A PEBBLE BED REACTOR

Description: A comprehensive, high fidelity model for pebble flow has been developed and embodied in the PEBBLES computer code. In this paper, a description of the physical artifacts included in the model is presented and some results from using the computer code for predicting the features of pebble flow and packing in a realistic pebble bed reactor design are shown. The sensitivity of models to various physical parameters is also discussed.
Date: October 1, 2006
Creator: Cogliati, Joshua J. & Ougouag, Abderrafi M.
Partner: UNT Libraries Government Documents Department

A comparison of the performance of compact neutrino detector designs for nuclear reactor safeguards and monitoring.

Description: There has been an increasing interest in the monitoring of nuclear fuel for power reactors by detecting the anti-neutrinos produced during operation. Small liquid scintillator detectors have already demonstrated sensitivity to operational power levels, but more sensitive monitoring requires improvements in the efficiency and uniformity of these detectors. In this work, we use a montecarlo simulation to investigate the detector performance of four different detector configurations. Based on the analysis of neutron detection efficiency and positron energy response, we find that the optimal detector design will depend on the goals and restrictions of the specific installation or application. We have attempted to present the relevant information so that future detector development can proceed in a profitable direction.
Date: October 27, 2006
Creator: Reyna, D. E.; McKeown, R. W.; Physics, High Energy & Univ., Drexel
Partner: UNT Libraries Government Documents Department

Enhancement of REBUS-3/DIF3D for whole-core neutronic analysis of prismatic very high temperature reactor (VHTR).

Description: Enhancements have been made to the REBUS-3/DIF3D code suite to facilitate its use for the design and analysis of prismatic Very High Temperature Reactors (VHTRs). A new cross section structure, using table-lookup, has been incorporated to account for cross section changes with burnup and fuel and moderator temperatures. For representing these cross section dependencies, three new modules have been developed using FORTRAN 90/95 object-oriented data structures and implemented within the REBUS-3 code system. These modules provide a cross section storage procedure, construct microscopic cross section data for all isotopes, and contain a single block of banded scattering data for efficient data management. Fission products other than I, Xe, Pm, and Sm, can be merged into a single lumped fission product to save storage space, memory, and computing time without sacrificing the REBUS-3 solution accuracy. A simple thermal-hydraulic (thermal-fluid) feedback model has been developed for prismatic VHTR cores and implemented in REBUS-3 for temperature feedback calculations. Axial conduction was neglected in the formulation because of its small magnitude compared to radial (planar) conduction. With the simple model, the average fuel and graphite temperatures are accurately estimated compared to reference STAR-CD results. The feedback module is currently operational for the non-equilibrium fuel cycle analysis option of REBUS-3. Future work should include the extension of this capability to the equilibrium cycle option of the code and additional verification of the feedback module. For the simulation of control rods in VHTR cores, macroscopic cross section deviations (deltas) have been defined to account for the effect of control rod insertion. The REBUS-3 code has been modified to use the appropriately revised cross sections when control rods are inserted in a calculation node. In order to represent asymmetric core blocks (e.g., fuel blocks or reflector blocks containing asymmetric absorber rods), surface-dependent discontinuity factors based on nodal ...
Date: October 13, 2006
Creator: Lee, C. H.; Zhong, Z.; Taiwo, T.A.; Yang, W.S.; Khalil, H.S.; Smith, M.A. et al.
Partner: UNT Libraries Government Documents Department