Clinch River Breeder Reactor Plant Research Articles

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Application of a Microprocessor Based Solid State Programmable Logic System for Control of Class 1E Components and Devices

The paper discusses the microprocessor based Solid State Programmable Logic System (SSPLS) designed for the Clinch River Breeder Reactor Plant (CRBRP) which was terminated in December 1983 by U.S. Congress. Even though procurement of equipment did not take place, the topics discussed here reflect the advanced engineering and design completion (85 percent) on the project and the authors feel that the information presented reflects worthwhile experience associated with the implementation of the state-of-the-art technology and may benefit future designs as well as the industry in general. Topics presented include discussion on design description and hardware, programming, verification and validation, licensing concerns and requirements on reliability and maintainability. The paper describes licensing interaction with the Nuclear Regulatory Commission (NRC) Staff at the construction permit stage and outlines additional details that would be needed at an operating license permit stage.

Elevated-temperature structural design evaluation issues in LMFBR licensing

This paper describes elevated-temperature structural design issues and concerns identified by the NRC licensing review of the Clinch River Breeder Reactor Plant for a construction permit. Major issues concern weldment evaluation, notch weakening, steam generator tubesheet evaluation, and the use of “average” rather than “minimum” material properties in inelastic analysis. Other questions concern seismic effects, elastic follow-up, creep fatigue evaluation, plastic strain concentration, and transition welds. All of the issues were resolved but several required CRBRP Project commitment to perform additional confirmatory programs.

THREE-DIMENSIONAL FINITE-ELEMENT ANALYSIS OF THE CELLULAR CONVECTION PHENOMENA IN THE CLINCH RIVER BREEDER REACTOR PLANT PROTOTYPE PUMP

Cellular convection was studied rigorously during the development of the Clinch River Breeder Reactor Plant (CRBRP) Program Pumps. The non-symmetrical effects of cellular convection on thermal stresses, thermal distortions, and pump/foundation interface temperatures were important considerations. This paper presents the development of a three-dimensional finite-element heat transfer model which accounts for the cellular convection phenomena. A buoyancy driven cellular convection flow pattern is introduced in the annulus region between the upper inner structure and the pump tank. Steady-state thermal data were obtained for several test conditions for argon gas pressures up to 93psig (741 kPa) and sodium operating temperatures to 1000°F (811 K). Test temperature distributions on the pump tank and inner structure were correlated with numerical results and excellent agreement was obtained.

Structural research needs for Clinch River Breeder Reactor licensing

This paper discusses several structural issues for the Clinch River Breeder Reactor Plant (CRBRP) that need to be investigated before the review for an operating license begins. The issues exist because of differences between the CRBRP structures and typical light water reactor (LWR) plant structures. CRBRP structures are generally larger and must withstand loads characterized by lower pressures and higher temperatures than occur in LWRs. In the paper each major issue is described along with research needed to determine its impact on plant safety.

Electrical Power Distribution System for Clinch River Breeder Reactor Plant

This paper describes the auxiliary electric power system for the Clinch River Breeder Reactor Plant (CRBRP). The power systems covered are the utility grid interconnections, plant power supply systems medium and low voltage AC systems, 125/250V DC system, 120V vital AC system, and safety related AC distribution system. The paper also describes the mode of connection of non-Class 1E loads to Class 1E power systems, and a cross-connect feature between Division 1 and Division 2 Class 1E AC Power Systems.

The Effect of Irradiation Induced Structural Material Deformations on Core Restraint Design

A summary of the major considerations in the design of LMFBR core restraint systems is presented. A discussion of these considerations is given using the design features and environment of the Clinch River Breeder Reactor Plant core restraint system design. A comparison of core restraint performance with an alternate concept is provided. Conclusions are drawn on the direction of current and future core restraint system designs.

Nuclear Fuel Cycle: Issues and Some Answers for the U.S.A.

The principal issues regarding nuclear power center on reprocessing and recycling of spent fuel, deployment of breeder reactors, and waste disposal. Analysis of future energy needs using a minimum societal expectation approach indicates that economically recoverable domestic uranium resources probably will have been exhausted early in the next century. Even if speculative resources are assumed to become available, deployment of commercial breeders may be required in the 2010-2020 time frame. Accordingly, a vigorous breeder and fuel recycle research, development and demonstration (RD&D) program is needed at this time. The plutonium needs for current breeder RD&D projects are a matter of some concern. The estimated requirements for the period 1987-1995 range from 12 to 30 tons of plutonium. A facility for producing U-Pu oxide fuel subassemblies for the Clinch River Breeder Reactor Plant must be operable by 1986 (there is no existing industrial capability for fabricating U-Pu fuels), and a pilot-scale breeder fuel reprocessing plant must be ready to accept CRBRP fuel in 1993. For waste disposal, an engineering-scale demonstration for high-level nuclear wastes is urgently needed.

Use of spar elements to stimulate fluid acoustical effects and fluid-solid interaction in the finite element analysis of piping system dynamics

A technique is presented for using existing options in structural analysis finite element computer programs to simulate fluid acoustical effects and fluid-structure interaction in piping system dynamic analysis. With this technique, the fluid in straight pipe section is represented as a sequence of spar elements coupled to the pipe motion in the transverse direction but free to move independently in the axial direction. Special modeling considerations for treating the acoustical and fluid-structure interaction effects at elbows, tees and area changes are also derived. Results using this kind of modeling are presented for a pulse loading of a leg of piping with the same cross-sectional dimensions and elbow radii as the Clinch River Breeder Reactor Plant (CRBRP) primary piping.

Dynamic Simulation of LMFBR Plant under Natural Circulation

Dynamic simulation of the entire LMFBR plant is needed to determine after-heat dissipation capability of the plant via natural circulation. Models required to accomplish this task are briefly described. The resulting computer code, Super System Code-Loop (SSC-L), is then applied to a loop-type fast reactor design to illustrate the system response for a loss-of-electric-power event. Where possible, the Clinch River Breeder Reactor Plant (CRBRP) specifications are used. Entire simulation is performed for up to 1800 seconds of the transient time. Even with rather detailed modeling of essential components, the computing time required on the CDC-7600 machine was slightly less than the real time. For the assumed plant conditions, it is found that the decay heat is adequately removed via natural circulation. The effect of advanced modeling on system performance is also demonstrated. Finally, a scenario of an event that could impose more stringent testing of the plant’s capability for adequate dissipation of decay heat is described.

Experimental investigations of the hydraulic field in wire-wrapped LMFBR core assemblies

Abstract An investigation of the hydraulic behavior of wire-wrapped fuel and blanket assemblies was conducted in an air flow test facility. The test section was a large scale sector (slightly more than one-sixth) of prototypic fuel and blanket assemblies of the Clinch River Breeder Reactor Plant; the scale factor was approximately 11:1 and 5:1 for the fuel and blanket, respectively, thus allowing a very large number of measurements within each subchannel. The purpose of these experiments is discussed along with a brief state of the art review; also discussed is the role of these tests on the core thermal-hydraulic design through calibration and verification of the analytical codes employed in the design. The test section and experimental procedures are illustrated. Experimental results are discussed in detail: static pressure gradients; local and average cross flow through the gap spacing between rods as a function of the wire wrap position and at all typical locations in the assembly; detailed axial velocity mappings in the inboard and peripheral channels. The physical significance of the results is interpreted and the fundamental difference in the hydraulic behavior of fuel and blanket assemblies is pointed out, discussed and explained in terms of fundamental geometric parameters. The application of the fuel assembly data to calibration/verification of subchannel analysis and distributed parameter codes is presented in detail. A quantitative model of the cross flow driving forces is elaborated as the starting point for a comprehensive phenomenological modeling of the hydraulic behavior of wire-wrapped assemblies.

Status and prospects of the clinch river breeder reactor plant project (CRBRP) entering the 1980s

Status and prospects of the clinch river breeder reactor plant project (CRBRP) entering the 1980s

Evolution of structural reliability methodology as applied in major nuclear risk assessments

The evolution of the handling of structure and mechanical component reliability in the seismic portions of five major nuclear risk assessments is reviewed. Three of these risk assessments are complete and published (the Reactor Safety Study, the Clinch River Breeder Reactor Plant Safety Study, and the Diable Canyon License Application Amendment 52), while the last two are still in progress (the Oyster Creek Probabilistic Safety analysis and the Seismic Safety Margins Research Program). The approaches and conclusions (to the extent available) are summarized and compared, demonstrating important advances in the sophistication of performing structural reliability analyses over a span of about eight years of major risk assessment effort.

Gas Tag Identification of Failed Reactor Assemblies—IV. Analysis Methods

An examination of tag gas designs for the Fast Flux Test Facility and the Clinch River Breeder Reactor Plant indicates that an unambiguous identification of an assembly with a breached pin should be possible under normal operating conditions when 1% or less of the tag gas escapes; correction for the background tag gas in the argon cover gas improves the probability of correct identification. The probabilistic method of Schmittroth has been further developed for use in situations in which the escaped tag gas does not permit a unique identification of a failed assembly.

Summary of the ALARA Nature of Design Effluents from the Clinch River Breeder Reactor Plant

Radioactive effluents and concentrations expected from the Clinch River Breeder Reactor (CRBR) during normal operation have been assessed. Developed were the individual off-site dose values as well as population doses within an 80-km (50-mile) radius of the proposed facility. Comparison has been with the as low as reasonably achievable (ALARA) exposure objective as in 10CFR50, Appendix I, of the Code of Federal Regulations. All estimated releases are well within the concentration limits found in 10CFR20. The calculated doses to individuals are consistently lower than values of 10CFR50, Appendix I. Population exposures are a very small fraction of that due to background radiation. Exposures due to CRBR effluents are also in conformity with cost-benefit arguments of the ALARA objective. Measures are in place to follow the design progression and operation of the facility to retain the ALARA aspect of the doses associated with the effluent releases.

Core Exit Instrumentation of the Clinch River Breeder Reactor Plant

Assembly exit thermocouples are chosen for the Clinch River Breeder Reactor Plant as the instrumentation providing the most useful information at the minimum cost. One thermocouple is positioned at...

Uniaxial compressive strengths of concrete for temperatures reaching 1033 K

A series of tests was conducted to evaluate effects of sustained heating on the strengths of limestone aggregate concrete for the Clinch River Breeder Reactor Plant. Twenty-four 15.24-cm dia. by 30.48-cm long cylinders were divided into two groups subjected to either open-hot or closed-cold conditions. All specimens were heated for 14 days at soak temperatures of 450, 644, 839 and 1033 K. Unconfined compressive strengths were determined at these temperatures for the open-hot cylinders and after cooling for the closed-cold set. The results provided realistic upper and lower bounds for published strength data.

Safety and Reliability- Analyses of a Breeder Reactor

This paper outlines a University of Tennessee indepen dent study of the safety and reliability of the Clinch River Breeder Reactor Plant in Oakridge, Tennessee. The study uses several kinds of models, and analytical computer pro grams, one of which (BACFIRE) is briefly described and illustrated.

Calculation of tube degradation induced by dryout instability in sodium-heated steam generators

Analytical methods are adapted and presented for the calculation of thermal fluctuations and thermal stresses experienced by a tube wall in the region of departure from nucleate boiling (DNB) in a sodium-heated steam generator. Calculated results are presented using parameter ranges and geometry adopted from the Atomics International reference design for the steam generators of the Clinch River Breeder Reactor Plant. The physical phenomenon was modeled by subjecting the tube wall and the oxide scale layer on the water side of the tube (treated as either a powdery substance or a dense protective film) to periodic water heat transfer coefficient fluctuations simulating the effect of DNB in that area of the tube. The stresses obtained were employed in estimating (where possible) the most detrimental contribution to the life expectancy of the steam generator tube based on fatigue of the tube wall and repeated exfoliation of the oxide layers.

Inelastic Behavior of Finite Circular Cylindrical Shells

In the design of elevated temperature components such as those encountered in Liquid Metal Fast Breeder Reactor (LMFBR) service, the designer/analyst is often faced with the task of having to assess structural adequacy of pressure vessel and piping components which experience high cyclic thermal stresses. Expensive and time consuming detailed inelastic analyses using finite element techniques are often necessary for such an assessment. Experience with the design of the LMFBR components has focused on an urgent need for simplified inelastic analysis methods which can aid the designer/analyst in scoping the design and minimize the number of parts requiring detailed inelastic analysis. Through its participation in the FFTF (Fast Flux Test Facility) and CRBRP (Clinch River Breeder Reactor Plant), Foster Wheeler Energy Corporation has developed a series of simplified analysis computer programs. The underlying philosophy in this work has been to make simplifying assumptions on the structural model but to solve the resulting boundary value problem as exactly as practicable so that approximations in the stress state or constitutive equations are not introduced. This paper is the third in a series [1, 11] by the authors dealing with the elastic-plastic-creep behavior of cylindrical structures. A rate formulation is presented for the elastic-plastic-creep analysis of finite circular cylindrical shells with various end conditions subjected to varying axisymmetric pressure loads, through-the-wall and along-the-length temperature gradients, and either axial loads or axial deformations. The solution procedure is based on direct integration and successive approximation and shown to be efficient in dealing with complicated loading histories. Applications of the present method of analysis are illustrated by numerical examples of elevated temperature design problem.