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Assessment of utilization of thorium in BWRs

Description: Performance characteristics and trends which could affect the incentives for and feasibility of use of thorium-based fuels in BWRs were investigated using preliminary scoping techniques. Benchmark comparisons with selected critical experiments and Monte Carlo calculations were made for simple geometries. Several fuel compositions, including uranium fissile isotopes diluted (''denatured'') with U-238 to reduce fissile enrichment below that suitable for weapon purposes, were evaluated. Emphasis was focused on design approaches which utilize unique BWR characteristics such as increased coolant boiling and nonuniform lattices. Thorium-based fuel material utilization efficiency in the BWR, without reprocessing and uranium recycle, shows no improvement in resource utilization relative to the UO/sub 2/ stowaway recycle. If plutonium recycle were restricted, then a major resource utilization incentive would exist for thorium use in LWRs with denatured uranium recycle and plutonium use in secured energy centers. The characteristics of hypothesized symbiotic systems involving secured plutonium burning sites are illustrated. However, the significant quantity of plutonium produced when ''denatured'' uranium is used raises question as to the nonproliferation effectiveness of the system. A distinctive incentive for thorium in the BWR appears to be its potential for favorable effects on power distribution, reactivity control, and margins for core dynamic response. The most promising designs are those which selectively locate thorium-bearing rods to most effectively utilize the unique heterogeneity of the BWR lattice, which is composed of individually channeled fuel assemblies separated by water spaces. Mixed lattice designs with only a few ThO/sub 2/ rods were found to significantly improve dynamic and control characteristics.
Date: January 1, 1978
Creator: Williamson, H.E.
Partner: UNT Libraries Government Documents Department

Identification of neutron noise sources in a boiling water reactor

Description: Measurements were made at Units 2 and 3 of the TVA Browns Ferry nuclear power plant in order to characterize the neutron and process signal noise signatures, to determine the degree of correlation between selected pairs of signals, and to assess the usefulness of such signatures for monitoring and anomaly detection in BWR-4s. Measurements were made in a power plant during normal operation at full power to determine the usefulness of the neutron and process signals from sensors and instrumentation in the plant which have been contaminated by plant electrical noise interference. It is concluded that the signals derived from existing plant sensors and instrumentation could be used to diagnose anomalies. The neutron signals could be used to monitor the stability of the core and to diagnose anomalies involving the reactor pressure, core flow, and steam flow.
Date: January 1, 1977
Creator: Sides, W.H. Jr. & Mathis, M.V.
Partner: UNT Libraries Government Documents Department

Numerical calculation of the global and local components of the neutron noise field in BWR's

Description: The calculation of the local and global components of the neutron noise field is illustrated. Agreement with experimental results is excellent above 6 H/sub 2/ frequency. The discrepancy at lower frequencies indicates the inadequacy of the point kinetics model to describe the global component of the neutron noise in large BWRs.
Date: January 1, 1979
Creator: Difilippo, F. C. & Otaduy, P. J.
Partner: UNT Libraries Government Documents Department

Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 1. Investigation and evaluation of stress corrosion cracking in piping of boiling water reactor plants

Description: IGSCC in BWR piping is occurring owing to a combination of material, environment, and stress factors, each of which can affect both the initiation of a stress-corrosion crack and the rate of its subsequent propagation. In evaluating long-term solutions to the problem, one needs to consider the effects of each of the proposed remedial actions. Mitigating actions to control IGSCC in BWR piping must be designed to alleviate one or more of the three synergistic factors: sensitized material, the convention BWR environment, and high tensile stresses. Because mitigating actions addressing each of these factors may not be fully effective under all anticipated operating conditions, mitigating actions should address two and preferably all three of the causative factors; e.g., material plus some control of water chemistry, or stress reversal plus controlled water chemistry.
Date: August 1, 1984
Partner: UNT Libraries Government Documents Department

Independent assessment of TRAC and RELAP5 codes through separate effects tests

Description: Independent assessment of TRAC-PF1 (Version 7.0), TRAC-BD1 (Version 12.0) and RELAP5/MOD1 (Cycle 14) that was initiated at BNL in FY 1982, has been completed in FY 1983. As in the previous years, emphasis at Brookhaven has been in simulating various separate-effects tests with these advanced codes and identifying the areas where further thermal-hydraulic modeling improvements are needed. The following six catetories of tests were simulated with the above codes: (1) critical flow tests (Moby-Dick nitrogen-water, BNL flashing flow, Marviken Test 24); (2) Counter-Current Flow Limiting (CCFL) tests (University of Houston, Dartmouth College single and parallel tube test); (3) level swell tests (G.E. large vessel test); (4) steam generator tests (B and W 19-tube model S.G. tests, FLECHT-SEASET U-tube S.G. tests); (5) natural circulation tests (FRIGG loop tests); and (6) post-CHF tests (Oak Ridge steady-state test).
Date: January 1, 1983
Creator: Saha, P.; Rohatgi, U.S.; Jo, J.H.; Neymotin, L.; Slovik, G.; Yuelys-Miksis, C. et al.
Partner: UNT Libraries Government Documents Department

Physical model of nonlinear noise with application to BWR stability

Description: Within the framework of the present model it is shown that the BWR reactor cannot be unstable in the linear sense, but rather it executes limited power oscillations of a magnitude that depends on the operating conditions. The onset of these oscillations can be diagnosed by the decrease in stochasticity in the power traces and by the appearance of harmonics in the PSD.
Date: January 1, 1983
Creator: March-Leuba, J. & Perez, R.B.
Partner: UNT Libraries Government Documents Department

Nonlinear dynamics of boiling water reactors

Description: Recent stability tests in Boiling Water Reactors (BWRs) have indicated that these reactors can exhibit the special nonlinear behavior of following a closed trajectory called limit cycle. The existence of a limit cycle corresponds to an oscillation of fixed amplitude and period. During these tests, such oscillations had their amplitudes limited to about +- 15% of the operating power. Since limit cycles are fairly insensitive to parameter variations, it is possible to operate a BWR under conditions that sustain a limit cycle (of fixed amplitude and period) over a finite range of reactor parameters.
Date: January 1, 1983
Creator: March-Leuba, J.; Cacuci, D.G. & Perez, R.B.
Partner: UNT Libraries Government Documents Department

Fuel utilization in a progressive conversion reactor (PCR)

Description: Preliminary studies indicate that for once-through fuel cycles, the PCR offers potential improvements over current LWRs in the following major areas: improved uranium utilization (reduced uranium demand), degraded plutonium product in spent fuel, reduced plutonium content of spent fuel, reduced amount of spent fuel, reduced fissile content of spent fuel, and reduced separative work.
Date: May 1, 1981
Creator: Leyse, C.F. & Judd, J.L.
Partner: UNT Libraries Government Documents Department

Limit cycles and bifurcations in nuclear systems

Description: This work provides a basis for scoping calculations to determine the dynamic behavior - both linear and nonlinear - of BWRs. Additional work is now underway to establish the feasibility of routine operation of nuclear systems in the nonlinear (limit-cycle) regime.
Date: January 1, 1986
Creator: Cacuci, D.G.; March-Leuba, J. & Perez, R.B.
Partner: UNT Libraries Government Documents Department

Fuel performance improvement program: description and characterization of HBWR Series H-2, H-3, and H-4 test rods

Description: The fabrication process and as-built characteristics of the HBWR Series H-2 and H-3 test rods, as well as the three packed-particle (sphere-pac) rods in HBWR Series H-4 are described. The HBWR Series H-2, H-3, and H-4 tests are part of the irradiation test program of the Fuel Performance Improvement Program. Fifteen rods were fabricated for the three test series. Rod designs include: (1) a reference dished pellet design incorporating chamfered edges, (2) a chamfered, annular pellet design combined with graphite-coated cladding, and (3) a sphere-pac design. Both the annular-coated and sphere-pac designs include internal pressurization using helium.
Date: March 1, 1980
Creator: Guenther, R.J.; Barner, J.O. & Welty, R.K.
Partner: UNT Libraries Government Documents Department

Fuel Performance Improvement Program. Quarterly progress report, April-June 1979

Description: Two series of test rods are under irradiation in the Halden Boiling Water Reactor (HBWR Series H-1 and Series H-4). Fuel rods for Series H-2 and H-3 have been fabricated and delivered to Halden. Plans for the first series of demonstration fuel assemblies for irradiation in the Big Rock Point Reactor have been modified to substitute two BRPR Series S-2 assemblies containing segmented rods, some with sphere-pac fuel, for Series S-1 assemblies that will be irradiated later. These assemblies are in the current BRPR reload and reactor startup is scheduled for September-October. An effective method for applying graphite coating to the inside surface of cladding tubes has been demonstrated. Rod loading procedures for sphere-pac fuel that assure a uniform axial fuel density have been developed. Special hardware has been designed and tested that assures the spherical fuel can be retained in the fuel column during loading and handling operations. The measured centerline fuel temperature in the pressurized sphere-pac fuel rod (HBWR Series H-4) is lower (approx. 145/sup 0/C) than for the reference pellet fuel rod operating at comparable linear heat ratings.
Date: July 1, 1979
Creator: Crouthamel, C.E. (comp.)
Partner: UNT Libraries Government Documents Department

Development of an automated diagnostic system for BWR stability measurements

Description: An algorithm capable of automatically evaluating BWR stability has been developed. Main advantages are: Conservative estimate (asymptotic), adjusts to solve difficult conditions, confidence level, and error estimate. The apparent decay ratio (DR) is not a conservative estimate of the reactor stability. The asymptotic decay ratio must be used. Long enough record lengths must be used to reduce the uncertainty of the estimated DR.
Date: October 1, 1984
Creator: March-Leuba, J. & Smith, C.M.
Partner: UNT Libraries Government Documents Department

Comparison of BIASI and Columbia CHF correlations using BODYFIT-2PE

Description: This paper compares the BIASI critical heat flux (CHF) correlation with the Columbia CHF correlation by using both the homogeneous equilibrium two-phase model with algebraic slip and the drift flux model in BODYFIT-2PE. All calculations were compared with the GE 3 x 3 CHF experiment. This comparison serves as a qualification process for the CHF correlations in the framework of BODYFIT-2PE.
Date: January 1, 1984
Creator: Chen, B. C. J.; Chien, T. H.; Sha, W. T. & Kim, J. H.
Partner: UNT Libraries Government Documents Department

Summary of ORNL investigation of in-core vibrations in BWR-4s

Description: This report describes the use of noise analysis to investigate in-core instrument tube vibrations in BWR-4 reactors. Neutron noise signals from in-core fission chambers and acoustic noise signals from externally mounted accelerometers were used in these studies. The results show that neutron noise can be used to detect vibration and, more importantly, impacting of instrument tubes against adjacent fuel channel boxes. Externally mounted accelerometers detect impacting but not rubbing of instrument tubes against fuel channel boxes. Accelerometers can monitor impacting only on the particular instrument tube where the accelerometer is mounted. Surveillance for instrument tube impacts can be accomplished using standard BWR-4 in-core power range neutron flux detectors at all instrument tube locations containing these detectors. Ex-vessel accelerometers can then be used to monitor instrument tubes that lack power range neutron flux detectors. However, noise on axial flux profiles obtained with movable in-core detectors is not a reliable indicator of impacting, because the recorder used to plot the flux profiles does not respond adequately to the noise frequency generated by impacting.
Date: March 25, 1977
Creator: Fry, D. N.; Kryter, R. C.; Mathis, M. V.; Mott, J. E. & Robinson, J. C.
Partner: UNT Libraries Government Documents Department

Existence of short and long range relaxation lengths in heterogeneous media

Description: Experimental data related to boiling water reactors show that the phase of the cross power spectra density between the detector response at two points in the system is described by a pure delay process above certain values of the frequency. For this to occur the adjoint flux (detector field of view) must be sharply peaked around each one of the detectors, which in turn implies the existence of a short range relaxation length. It is shown that the existence of this type of relaxation length is a direct consequence of the physical properties of heterogeneous systems and not a consequence of the number of groups used to describe the neutron field in the equivalent homogeneous system as it has been done traditionally. To show this point a one-group neutron diffusion model which accounts explicitly for the heterogeneities of the system is presented.
Date: January 1, 1980
Creator: Difilippo, F.C.
Partner: UNT Libraries Government Documents Department

NASAP: a computer code for the evaluation of the Non-proliferation Alternative Systems Assessment Program concepts. Final report in support of Task 2. [PWR; BWR]

Description: The Non-Proliferation Alternative Systems Assessment Program (NASAP) computer code was developed to calculate the LWR and NASAP choice reactor cost through an arbitrary year T/sub N/. The final cost is arrived at by calculation of cost contributory factors for both LWR and NASAP choice reactors.
Date: September 1, 1979
Creator: Maul, B. A.
Partner: UNT Libraries Government Documents Department

Fuel performance improvement program. Quarterly/annual progress report, October 1977--September 1978. [BWR; PWR]

Description: This quarterly/annual report reviews and summarizes the activities performed in support of the Fuel Performance Improvement Program (FPIP) during Fiscal Year 1978 with emphasis on those activities that transpired during the quarter ending September 30, 1978. Significant progress has been made in achieving the primary objectives of the program, i.e., to demonstrate commercially viable fuel concepts with improved fuel - cladding interaction (FCI) behavior. This includes out-of-reactor experiments to support the fuel concepts being evaluated, initiation of instrumented test rod experiments in the Halden Boiling Water Reactor (HBWR), and fabrication of the first series of demonstration rods for irradiation in the Big Rock Point Reactor (BRPR).
Date: October 1, 1978
Creator: Crouthamel, C.E. (comp.)
Partner: UNT Libraries Government Documents Department

Effect of non-heterogeneous wetwell boundaries on pressure suppression system response. [BWR]

Description: The Full-Scale Mark II CRT (Containment Response Test) Program is in progress at the Tokai-Mura Establishment of the Japan Atomic Energy Research Institute (JAERI). The primary objective of the on-going CRT Program is to provide a data base for evaluation of the pressure suppression pool (wetwell) hydrodynamic loads associated with a postulated loss-of-coolant accident (LOCA) in the BWR Mark II containment system. The test facility is 1/18 of full scale in volume and has a wetwell which is a full-scale geometric replica of one 20/sup 0/-sector of a reference 1100MWe Mark II.
Date: August 29, 1980
Creator: McCauley, E.W.; Holman, G.S.; Namatame, K.; Kukita, Y. & Shiba, M.
Partner: UNT Libraries Government Documents Department

Development of BWR (boiling water reactor) and PWR (pressurized water reactor) event descriptions for nuclear facility simulator training

Description: A number of tools that can aid nuclear facility training developers in designing realistic simulator scenarios have been developed. This paper describes each of the tools, i.e., event lists, events-by-competencies matrices, and event descriptions, and illustrates how the tools can be used to construct scenarios.
Date: January 1, 1987
Creator: Carter, R.J. & Bovell, C.R.
Partner: UNT Libraries Government Documents Department

Aging and Service Wear of Air Compressors and Dryers in Nuclear Power Plants

Description: Compressed air systems and their associated compressors and dryers as incorporated in LWR power plants usually are not classified as safety related systems and components because their continued operation is not required to bring the plant to a safe shutdown condition. However, control air is a vital requirement for maintaining stable plant operation and its loss often results in a reactor trip and, on occasion, the actuation of engineered safety feature systems. Concerns caused by repeated instances of failures in air systems resulted in the high-priority ranking of Generic Issue No. 43, ''Reliability of Air Systems'' by the NRC. In support of the Nuclear Plant Aging Research (NPAR) Program, the Oak Ridge National Laboratory is carrying out a Phase I aging assessment of air compressors and dryers used in LWR power plants. The objectives of this study include: (1) determination of the types and ratings of equipment utilized in typical plants; (2) identification of aging and service wear stressors imposed on the machines; (3) evaluation of operating experience with the machines; (4) comparison of user and manufacturer-recommended inspection, surveillance, and monitoring (ISM) methods; and (5) the identification of any improved ISM methods that might lead to longer, more reliable, service. 2 refs.
Date: January 1, 1989
Creator: Moyers, J. C.
Partner: UNT Libraries Government Documents Department

Advanced Light Water Reactor Program: Program management and staff review methodology

Description: This report summarizes the NRC/EPRI coordinated effort to develop design requirements for a standardized advanced light water reactor (ALWR) and the procedures for screening and applying new generic safety issues to this program. The end-product will be an NRC-approved ALWR Requirements Document for use by the nuclear industry in generating designs of LWRs to be constructed for operation in the 1990s and beyond.
Date: December 1, 1986
Creator: Moran, D.H.
Partner: UNT Libraries Government Documents Department

Analysis of nuclear reactor instability phenomena

Description: The phenomena known as density-wave instability often occurs in phase change systems, such as boiling water nuclear reactors (BWRS). Our current understanding of density-wave oscillations is in fairly good shape for linear phenomena (eg, the onset of instabilities) but is not very advanced for non-linear phenomena [Lahey and Podowski, 1989]. In particular, limit cycle and chaotic instability modes are not well understood in boiling systems such as current and advanced generation BWRs (eg, SBWR). In particular, the SBWR relies on natural circulation and is thus inherently prone to problems with density-wave instabilities. The purpose of this research is to develop a quantitative understanding of nonlinear nuclear-coupled density-wave instability phenomena in BWRS. This research builds on the work of Achard et al [1985] and Clausse et al [1991] who showed, respectively, that Hopf bifurcations and chaotic oscillations may occur in boiling systems.
Date: January 1, 1993
Creator: Lahey, R.T. Jr.
Partner: UNT Libraries Government Documents Department

Radiation level assessment of the Dresden-1 decontamination pilot loop

Description: The radionuclide concentrations of the Dresden-1 decontamination pilot loop were determined by gamma spectroscopy. The General Electric Ge(Li)pipe gamma scanning system was utilized to take measurements at eight locations both before and after the pilot demonstration of decontamination process. Dose rate measrurements were taken with a portable gamma monitor at 30 additional locations. The percentage of Co-60 removed was calculated and the results were interpreted.
Date: May 1, 1978
Partner: UNT Libraries Government Documents Department