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DOE-EPRI On-Line Monitoring Implementation Guidelines

Description: Industry and EPRI experience at several plants has shown on-line monitoring to be very effective in identifying out-of-calibration instrument channels or indications of equipment-degradation problems. The EPRI implementation project for on-line monitoring has demonstrated the feasability of on-line monitoring at several participating nuclear plants. The results have been very enouraging, and substantial progress is anticipated in the coming years.
Date: January 2, 2003
Creator: E. Davis, R. Bickford
Partner: UNT Libraries Government Documents Department

Isotopic Generation and Confirmation of the PWR Application Model

Description: The objective of this calculation is to establish an isotopic database to represent commercial spent nuclear fuel (CSNF) from pressurized water reactors (PWRs) in criticality analyses performed for the proposed Monitored Geologic Repository at Yucca Mountain, Nevada. Confirmation of the conservatism with respect to criticality in the isotopic concentration values represented by this isotopic database is performed as described in Section of the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000). The isotopic database consists of the set of 14 actinides and 15 fission products presented in Section of YMP 2000 for use in CSNF burnup credit. This set of 29 isotopes is referred to as the principal isotopes. The oxygen isotope from the UO{sub 2} fuel is also included in the database. The isotopic database covers enrichments of {sup 235}U ranging from 1.5 to 5.5 weight percent (wt%) and burnups ranging from approximately zero to 75 GWd per metric ton of uranium (mtU). The choice of fuel assembly and operating history values used in generating the isotopic database are provided is Section 5. Tables of isotopic concentrations for the 29 principal isotopes (plus oxygen) as a function of enrichment and burnup are provided in Section 6.1. Results of the confirmation of the conservatism with respect to criticality in the isotopic concentration values are provided in Section 6.2.
Date: November 10, 2003
Creator: Wimmer, L.B.
Partner: UNT Libraries Government Documents Department

U.S. Nuclear Power Plant Operating Cost and Experience Summaries

Description: The ''U.S. Nuclear Power Plant Operating Cost and Experience Summaries'' (NUREG/CR-6577, Supp. 2) report has been prepared to provide historical operating cost and experience information on U.S. commercial nuclear power plants during 2000-2001. Costs incurred after initial construction are characterized as annual production costs, which represent fuel and plant operating and maintenance expenses, and capital expenditures related to facility additions/modifications, which are included in the plant capital asset base. As discussed in the report, annual data for these two cost categories were obtained from publicly available reports and must be accepted as having different degrees of accuracy and completeness. Treatment of inconclusive and incomplete data is discussed. As an aid to understanding the fluctuations in the cost histories, operations summaries for each nuclear unit are provided. The intent of these summaries is to identify important operating events; refueling, major maintenance, and other significant outages; operating milestones; and significant licensing or enforcement actions. Information used in the summaries is condensed from operating reports submitted by the licensees, the Nuclear Regulatory Commission (NRC) database for enforcement actions, and outage reports.
Date: September 18, 2003
Creator: Reid, RL
Partner: UNT Libraries Government Documents Department


Description: This report describes the detailed results for task 2 of DOE-NERI project number 99-119 entitled ''Automatic Development of Highly Reliable Control Architecture for Future Nuclear Power Plants''. This project is a collaboration effort between the Oak Ridge National Laboratory (ORNL,) The University of Tennessee, Knoxville (UTK) and the North Carolina State University (NCSU). UTK is the lead organization for Task 2 under contract number DE-FG03-99SF21906. Under task 2 we completed the development of data-driven models for the characterization of sub-system dynamics for predicting state variables, control functions, and expected control actions. We have also developed the ''Principal Component Analysis (PCA)'' approach for mapping system measurements, and a nonlinear system modeling approach called the ''Group Method of Data Handling (GMDH)'' with rational functions, and includes temporal data information for transient characterization. The majority of the results are presented in detailed reports for Phases 1 through 3 of our research, which are attached to this report.
Date: April 10, 2003
Creator: Upadhyaya, B.R.
Partner: UNT Libraries Government Documents Department


Description: This report covers the first year of this three-year research grant under the University Coal Research program. The overall objective of this project is to develop a comprehensive kinetic model for slurry phase Fischer-Tropsch synthesis on iron catalysts. This model will be validated with experimental data obtained in a stirred tank slurry reactor (STSR) over a wide range of process conditions. The model will be able to predict concentrations of all reactants and major product species (H{sup 2}O, CO{sub 2}, linear 1- and 2-olefins, and linear paraffins) as a function of reaction conditions in the STSR. During the reporting period we have completed one STSR test with precipitated iron catalyst obtained from Ruhrchemie AG (Oberhausen-Holten, Germany). This catalyst was initially in commercial fixed bed reactors at Sasol in South Africa. The catalyst was tested at 13 different sets of process conditions, and had experienced a moderate deactivation during the first 500 h of testing (decrease in conversion from 56% to 50% at baseline process conditions). The second STSR test has been initiated and after 270 h on stream, the catalyst was tested at 6 different sets of process conditions.
Date: September 29, 2003
Creator: Bukur, Dragomir B.; Froment, Gilbert F.; Nowicki, Lech; Wang, Jiang & Ma, Wen-Ping
Partner: UNT Libraries Government Documents Department

Autonomous Control of Nuclear Power Plants

Description: A nuclear reactor is a complex system that requires highly sophisticated controllers to ensure that desired performance and safety can be achieved and maintained during its operations. Higher-demanding operational requirements such as reliability, lower environmental impacts, and improved performance under adverse conditions in nuclear power plants, coupled with the complexity and uncertainty of the models, necessitate the use of an increased level of autonomy in the control methods. In the opinion of many researchers, the tasks involved during nuclear reactor design and operation (e.g., design optimization, transient diagnosis, and core reload optimization) involve important human cognition and decisions that may be more easily achieved with intelligent methods such as expert systems, fuzzy logic, neural networks, and genetic algorithms. Many experts in the field of control systems share the idea that a higher degree of autonomy in control of complex systems such as nuclear plants is more easily achievable through the integration of conventional control systems and the intelligent components. Researchers have investigated the feasibility of the integration of fuzzy logic, neural networks, genetic algorithms, and expert systems with the conventional control methods to achieve higher degrees of autonomy in different aspects of reactor operations such as reactor startup, shutdown in emergency situations, fault detection and diagnosis, nuclear reactor alarm processing and diagnosis, and reactor load-following operations, to name a few. With the advancement of new technologies and computing power, it is feasible to automate most of the nuclear reactor control and operation, which will result in increased safety and economical benefits. This study surveys current status, practices, and recent advances made towards developing autonomous control systems for nuclear reactors.
Date: October 20, 2003
Creator: Basher, H.
Partner: UNT Libraries Government Documents Department

University Programs of the U.S. Advanced Fuel Cycle Initiative

Description: As the Advanced Accelerator Applications (AAA) Program, which was initiated in fiscal year 2001 (FY01), grows and transitions to the Advanced Fuel Cycle (AFC) Program in FY03, research for its underlying science and technology will require an ever larger cadre of educated scientists and trained technicians. In addition, other applications of nuclear science and engineering (e.g., proliferation monitoring and defense, nuclear medicine, safety regulation, industrial processes, and many others) require increased academic and national infrastructure and even larger student populations. Because of the recognition of these current and increasing requirements, the DOE began a multi-year program to involve university faculty and students in various phases of these Projects to support the infrastructure requirements of nuclear energy, science and technology fields as well as the special needs of the DOE transmutation program. Herein I summarize the goals and accomplishments of the university programs that have supported the AAA and AFC Programs during FY02, including the involvement of 120 students at more than 30 universities in the U.S. and abroad. I also highlight contributions to academic research from LANL, which hosted students from and sponsored research at more than 18 universities by more than 50 students and 20 faculty members, investing about 10% of its AFC budget.
Date: January 1, 2003
Creator: Beller, D. E. (Denis E.)
Partner: UNT Libraries Government Documents Department

The Addition of Noncondensable Gases into RELAP5-3D for Analysis of High Temperature Gas-Cooled Reactors

Description: Oxygen, carbon dioxide, and carbon monoxide have been added to the RELAP5-3D computer code as noncondensable gases to support analysis of high temperature gas-cooled reactors. Models of these gases are required to simulate the effects of air ingress on graphite oxidation following a loss-of-coolant accident. Correlations were developed for specific internal energy, thermal conductivity, and viscosity for each gas at temperatures up to 3000 K. The existing model for internal energy (a quadratic function of temperature) was not sufficiently accurate at these high temperatures and was replaced by a more general, fourth-order polynomial. The maximum deviation between the correlations and the underlying data was 2.2% for the specific internal energy and 7% for the specific heat capacity at constant volume. The maximum deviation in the transport properties was 4% for oxygen and carbon monoxide and 12% for carbon dioxide.
Date: August 1, 2003
Creator: Davis, C. B. & Oh, C. H.
Partner: UNT Libraries Government Documents Department

NGNP Point Design - Results of the Initial Neutronics and Thermal-Hydraulic Assessments During FY-03, Rev. 1

Description: This report presents the preliminary preconceptual designs for two possible versions of the Next Generation Nuclear Plant (NGNP), one for a prismatic fuel type helium gas-cooled reactor and one for a pebble bed fuel helium gas reactor. Both designs are to meet three basic requirements: a coolant outlet temperature of 1000 °C, passive safety, and a total power output consistent with that expected for commercial high-temperature gas-cooled reactors. The two efforts are discussed separately below. The analytical results presented in this report are very promising, however, we wish to caution the reader that future, more detailed, design work will be needed to provide final answers to a number of key questions including the allowable power level, the inlet temperature, the power density, the optimum fuel form, and others. The point design work presented in this report provides a starting point for other evaluations, and directions for the detailed design, but not final answers.
Date: September 1, 2003
Creator: MacDonald, Philip E.; Sterbentz, James W.; Sant, Robert L.; Bayless, P.; Gougar, H. D.; Moore, R. L. et al.
Partner: UNT Libraries Government Documents Department

The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

Description: The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.
Date: September 1, 2003
Creator: Weaver, K. D.; Marshall, T.; Wei, T. Y. C.; Feldman, E. E.; Driscoll, M. J. & Ludewig, H.
Partner: UNT Libraries Government Documents Department

Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

Description: The regulatory requirement to develop an upgraded safety basis for a DOE nuclear facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830).1 Subpart B of 10 CFR 830, “Safety Basis Requirements,” requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements.1 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants”2 as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.
Date: June 1, 2003
Creator: Sharp, Gregg L. & McCracken, R. T.
Partner: UNT Libraries Government Documents Department

Reactor Physics Parametric and Depletion Studies in Support of TRISO Particle Fuel Specification for the Next Generation Nuclear Plant

Description: Reactor physics calculations were initiated to answer several major questions related to the proposed TRISO-coated particle fuel that is to be used in the prismatic Very High Temperature Reactor (VHTR) or the Next Generation Nuclear Plant (NGNP). These preliminary design evaluation calculations help ensure that the upcoming fuel irradiation tests will test appropriate size and type of fuel particles for a future NGNP reactor design. Conclusions from these calculations are expected to confirm and suggest possible modifications to the current particle fuel parameters specified in the evolving Fuel Specification. Calculated results dispel the need for a binary fuel particle system, which is proposed in the General Atomics GT-MHR concept. The GT-MHR binary system is composed of both a fissile and fertile particle with 350- and 500- micron kernel diameters, respectively. For the NGNP reactor, a single fissile particle system (single UCO kernel size) can meet the reactivity and power cycle length requirements demanded of the NGNP. At the same time, it will provide substantial programmatic cost savings by eliminating the need for dual particle fabrication process lines and dual fuel particle irradiation tests required of a binary system. Use of a larger 425-micron kernel diameter single fissile particle (proposed here), as opposed to the 350-micron GT-MHR fissile particle size, helps alleviate current compact particle packing fractions fabrication limitations (<35%), improves fuel block loading for higher n-batch reload options, and tracks the historical correlation between particle size and enrichment (10 and 14 wt% U-235 particle enrichments are proposed for the NGNP). Overall, the use of the slightly larger kernel significantly broadens the NGNP reactor core design envelope and provides increased design margin to accommodate the (as yet) unknown final NGNP reactor design. Maximum power-peaking factors are calculated for both the initial and equilibrium NGNP cores. Radial power-peaking can be fully controlled ...
Date: September 1, 2003
Creator: Sterbentz, James W.; Phillips, Bren; Sant, Robert L.; Chang, Gray S. & Bayless, Paul D.
Partner: UNT Libraries Government Documents Department

Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability

Description: Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with ...
Date: September 1, 2003
Creator: MacDonald, Philip E.
Partner: UNT Libraries Government Documents Department

Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

Description: This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.
Date: April 1, 2003
Creator: Kingrey, Karel I.
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

New Generation Nuclear Plant -- High Level Functions and Requirements

Description: This functions and requirements (F&R) document was prepared for the Next Generation Nuclear Plant (NGNP) Project. The highest-level functions and requirements for the NGNP preconceptual design are identified in this document, which establishes performance definitions for what the NGNP will achieve. NGNP designs will be developed based on these requirements by commercial vendor(s).
Date: September 1, 2003
Creator: Ryskamp, J. M.; Gorski, E. J.; Harvego, E. A.; Khericha, S. T. & Beitel, G. A.
Partner: UNT Libraries Government Documents Department

Multi-Applications Small Light Water Reactor - NERI Final Report

Description: The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.
Date: December 1, 2003
Creator: Modro, S. Michale; Fisher, James E.; Weaver, Kevan D.; Jose N. Reyes, Jr.; Groome, John T.; Babka, Pierre et al.
Partner: UNT Libraries Government Documents Department

Continuous-wave radar to detect defects within heat exchangers and steam generator tubes.

Description: A major cause of failures in heat exchangers and steam generators in nuclear power plants is degradation of the tubes within them. The tube failure is often caused by the development of cracks that begin on the outer surface of the tube and propagate both inwards and laterally. A new technique was researched for detection of defects using a continuous-wave radar method within metal tubing. The experimental program resulted in a completed product development schedule and the design of an experimental apparatus for studying handling of the probe and data acquisition. These tests were completed as far as the prototypical probe performance allowed. The prototype probe design did not have sufficient sensitivity to detect a defect signal using the defined radar technique and did not allow successful completion of all of the project milestones. The best results from the prototype probe could not detect a tube defect using the radar principle. Though a more precision probe may be possible, the cost of design and construction was beyond the scope of the project. This report describes the probe development and the status of the design at the termination of the project.
Date: January 1, 2003
Creator: Nassersharif, Bahram (New Mexico State University, Las Cruces, NM); Caffey, Thurlow Washburn Howell; Jedlicka, Russell P. (New Mexico State University, Las Cruces, NM); Garcia, Gabe V. (New Mexico State University, Las Cruces, NM) & Rochau, Gary Eugene
Partner: UNT Libraries Government Documents Department

RELAP-5/MOD 3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

Description: The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior.
Date: January 1, 2003
Creator: Bayless, Paul D.
Partner: UNT Libraries Government Documents Department

A TRAC model of the Los Alamos National Laboratory DELTA loop facility

Description: Lead-bismuth eutectic (LBE) is being used as one of the prospective coolants in advanced reactors and accelerator-driven systems because of its attractive nuclear, thermal, physical, and chemical properties. However, LBE is very corrosive to many structure materials if they are exposed to LBE directly. Thus, one of the critical obstacles to the wide use of LBE as a nuclear coolant is corrosion. A materials test loop facility, the DELTA Loop, has been set up at Los Alamos National Laboratory (LANL) to study the long-term corrosion of various materials in flowing LBE and to study the thermal hydraulics of such systems.
Date: January 1, 2003
Creator: Lime, J. F. (James F.) & Zhang, J. (Jinsuo)
Partner: UNT Libraries Government Documents Department

Reactor Physics Studies of Reduced-Tantaulum-Content Control and Safety Elements for the High Flux Isotope Reactor

Description: Some of the unirradiated High Flux Isotope Reactor (HFIR) control elements discharged during the late 1990s were observed to have cladding damage--local swelling or blistering. The cladding damage was limited to the tantalum/europium interface of the element and is thought to result from interaction of hydrogen and europium to form a compound of lower density than europium oxide, thus leading to a ''blistering'' of the control plate cladding. Reducing the tantalum loading in the control plates should help preclude this phenomena. The impact of the change to the control plates on the operation of the reactor was assessed. Regarding nominal, steady-state reactor operation, the impact of the change in the power distribution in the core due to reduced tantalum content was calculated and found to be insignificant. The magnitude and impact of the change in differential control element worth was calculated, and the differential worths of reduced tantalum elements vs the current elements from equivalent-burnup critical configurations were determined to be unchanged within the accuracy of the computational method and relevant experimental measurements. The location of the critical control elements symmetric positions for reduced tantalum elements was found to be 1/3 in. less withdrawn relative to existing control elements regardless of the value of fuel cycle burnup (time in the fuel cycle). The magnitude and impact of the change in the shutdown margin (integral rod worth) was assessed and found to be unchanged. Differential safety element worth values for the reduced-tantalum-content elements were calculated for postulated accident conditions and were found to be greater than values currently assumed in HFIR safety analyses.
Date: November 1, 2003
Creator: Primm, R.T., III
Partner: UNT Libraries Government Documents Department

A Model for Predicting Fission Product Activities in Reactor Coolant: Application of Model for Estimating I-129 Levels in Radioactive Waste

Description: A general model was developed to estimate the activities of fission products in reactor coolant and hence to predict a value for the I-129/Cs-137 scaling factor; the latter can be applied along with measured Cs-137 activities to estimate I-129 levels in reactor waste. The model accounts for fission product release from both defective fuel rods and uranium contamination present on in-core reactor surfaces. For simplicity, only the key release mechanisms were modeled. A mass balance, considering the two fuel source terms and a loss term due to coolant cleanup was solved to estimate fission product activity in the primary heat transport system coolant. Steady state assumptions were made to solve for the activity of shortlived fission products. Solutions for long-lived fission products are time-dependent. Data for short-lived radioiodines I-131, I-132, I-133, I-134 and I-135 were analyzed to estimate model parameters for I-129. The estimated parameter values were then used to determine I-1 29 coolant activities. Because of the chemical affinity between iodine and cesium, estimates of Cs-137 coolant concentrations were also based on parameter values similar to those for the radioiodines; this assumption was tested by comparing measured and predicted Cs-137 coolant concentrations. Application of the derived model to Douglas Point and Darlington Nuclear Generating Station plant data yielded estimates for I-129/I-131 and I-129/Cs-137 which are consistent with values reported for pressurized water reactors (PWRs) and boiling water reactors (BWRs). The estimated magnitude for the I-129/Cs-137 ratio was 10-8 - 10-7.
Date: February 27, 2003
Creator: Lewis, B. J. & Husain, A.
Partner: UNT Libraries Government Documents Department

Ion Chamber Compensation Tests

Description: The purpose of this report is to present the results of a series of tests performed to determine the need for gamma compensation of the ion chambers used to monitor the neutron flux in the 100 Area reactors.
Date: February 14, 2003
Creator: Mallard, R.L. Jr.
Partner: UNT Libraries Government Documents Department