Study of nuclear power plants incorporating the Spectral Shift Control Reactor (SSCR).

The objectives for this study were to determine fuel fabrication and assembly costs in a large nuclear power industry for several reactors, fuel rod sizes, and fuel materials.

From abstract: "Post-irradiation examination consisted of dimensional measurements, gamma scans, determination of fission gas release, visual examination of the fuel, measurement central voids, and metallographic examination of selected samples.

Report documenting research and developments made by the Health Physics Division of the Oak Ridge National Laboratory.

Fourth part of the "final report of a study directed toward the evolution, design, and demonstration of the principle design features of interim indirect cycle water cooled and moderated nuclear power plants which will be useful in early cooperative programs between the Atomic Energy Commission and the United States maritime industry" (p. i).

Abstract: This technical report elucidates principles of hazards evaluation. The concept of hazards potential is introduced and utilized to show how a reactor system perturbation will influence its nuclear safety. Literature relating to reactor safety is referenced to provide the sources of information required for hazards analysis and show how they influence a hazards evaluation. A checklist of items which should be considered in evaluating a change, test, or modification is presented.

Report describing the status of the SM-1, SM-1A, and PM-2A reactors, specifically regarding the effects "of irradiation on nil-ductility transition temperature and the associated problem of brittle fracture." (p. iii)

Abstract: The removal of both SM-1 Core I high burnup elements from the SM-1 Core II and the insertion of two SM-1 Core I spare elements i their places are discussed. Nuclear and thermal characteristics of Core II with the change are presented and conclusion related to the change in hazard potential are made. If the core change indicated by this report is made, local peaking factors will be decreased and burnout ratios will be increased. This, of course, in itself leads to a more conservative estimate of core safety. There is no conceivable reason why this perturbation may not be safely made in the SM-1 Core II.

Abstract: The removal of both SM-1 Core I high burnup elements from SM-1 Core II and the insertion of the PM-1-M-2 element and the SM-1 Core I spare element in SM-1 Core II is discussed. Nuclear and thermal characteristics of Core II with these changes are presented and conclusions related to the changes in the hazard potential are made. If the core change indicated by this report is made, local peaking factors will be decreased and burnout ratios will be increased. This, of course, in itself leads to a more conservative estimate of core safety. There is no conceivable reason why the perturbation may not be safely made in the SM-1 Core II.

Abstract: This technical report presents design criteria to prevent the brittle fracture of ferritic reactor vessels that cold occur as a result of the rise in NDT caused by fast neutron irradiation. The criteria require that maximum principal stress in the vessel does not exceed 18 percent of yield stress at temperatures below NDT + 60 degree F. Under certain conditions the allowable stress may be based on the irradiated yield stress. A discussion of brittle fracture and an explanation of the criteria are included.

Report documenting the ongoing research and developments of the Chemical Technology Division of the Oak Ridge National Laboratory.

Abstract: This technical report describes the as-constructed SM-1 penetrated gasket designed for SM-1 Core and Flow Instrumentation (Task XIV). This report supplements APAE No. 79, The Summary Hazards Report for Task XIV, and evaluates the effects of a postulated failure of this gasket. The effects of failure on the Maximum Credible Accident are determined and conclusions and recommendations for the use of this gasket are made.

Abstract: Thermal characteristics of the SM-2 core were analyzed at steady state and loss of flow conditions. For steady state operation, the steady state code STDY-3 was used. For transients during a loss of flow accident, ART-02, a one-dimensional code, was used. This analysis indicates the SM-2 core is safe from burnout under steady state operation at design power level (28 tMW) because (1) no nucleate boiling exists, and (2) the minimum burnout ratio is above 2.0. The core is safe from burnout under loss of flow transient because the minimum burnout ratio in the hottest element channel of 1. 82 is above the minimum design criteria of 1. 5.

From abstract: "The design, fabrication, and irradiation results are described for a 0.028 inch thick 304 stainless clad fuel element (SH-4B) exposed in the Vallecitos Boiling Water Reactor loop under superheat conditions."

Report describing tests "in which a double chambered irradiation capsule containing NaK (sodium-potassium alloy) in its inner chamber was deliberately perforated to allow NaK (the heat transfer medium) and water to react within the capsule chambers" (p. 2). The report includes descriptions of the materials used in the tests, test procedures, and results.

Abstract: This technical report describes the Martin Co. PM-1-M-2 test element and analyzes the potential hazard incurred by its inclusion in the SM-1 Core II. A nuclear analysis develops power distributions and reactivity effects. Hydraulic and thermal analyses develop anticipated burnout heat flux ratios. An evaluation of the risk involved with the inclusion of this element is presented. In view of the narrow margin by which the PM-1-M-2 test element meets the minimum burnout ratios as defined by Alco Products, Inc., it is recommended that if time permits that critical facility design verification be accomplished. The PM-1-M-2 test element meets the minimum requirements for insertion in SM-1 Core II and in view of the importance of this element to the PM-1 and PM-3A program, should be considered for insertion.

Abstract: Hydrodynamic flow studies were conducted on a full scale model of the SM-2 reactor vessel and core. Test fluid was water at 200 psi and 200 degree F. Test facilities, model, and instrumentation design are discussed. Flow distribution in the stationary fuel elements, lattices, and control rods of the second pass was investigated. Pressure losses through the various core components were measured and are compared with calculated values. Observed over-all pressure drop was 71 feet of water at 200 degree F, 31% higher than predicted, part of which was due to presence of instrument leads. Element to element flow distribution varied approximately +-8% from pass average. Channel-to-channel stationary element flow distribution varied approximately +-10% from element average and control rod flow distribution varied from +-8.9% to +-6.4 and -11.6% depending upon rod locations. These variations exceed the original goals of a +-10% and +-12% combined deviation for stationary and control rod elements respectively, but are satisfactory in relation to thermal design. There was no indication of unsatisfactory structural performance of any components under hydrodynamic loadings up to 130% of design values. The test program was terminated after determining flow distribution in the reference core design, omitting any work on …

From introduction: "A recent report describes the development of a general program for the IBM Type 7090 electronic computer for calculating plane thermal stresses."

From introduction: "The purpose of this report is to present the details of the exploratory and developmental work on tandem rolling, and the subsequent fabrication of tandem rolled fuel rods for irradiation testing in the Vallecitos Boiling Water Reactor."

Abstract: This technical report contains a description and results of a second series of critical experiments performed on the SM-2 core mock-up, as additional to the first series of experiments reported in APAE No. 54. The SM-2 core mock-up contains 36.4 kg U-235 and and estimated 67.9 gm B-10. The equivalent diameter and the active height are about 22 in.; the metal-to-water volume ration is 0.344. Data is presented on activation, reactivity, and stuck rod measurements. All measurements were conducted on the open seven control rod array employing 38 stationary fuel elements. Activation measurements consisted of neutron flux measurements using uranium fission foils for relative power distribution studies, the effect of flux suppressors on reducing power peaks, blocked coolant channel measurements, and gamma ray dose distribution. Reactivity measurements were performed to determine the effect f flow divider, flux suppressors and stimulated high temperature and pressure operation; b-10 loading in the SM-2 core; and core material coefficients. For the later, the worth in cents per gm or cents per cc was determined at simulated temperature of 510 degree F for B-10, U-235, stainless steel, and void. Stuck rod measurements were made to obtain an indication of the criticality margin in the …

Report containing ongoing projects and experiments undertaken by the Oak Ridge National Laboratory's Molten-Salt Reactor Program.

This report consists of the description, drawings, connections, and schematics of the various control elements that make up the control system of the Engineering Test Reactor Critical Facility (ETRC).

Abstract; SM-1 Reactor Core I Rearranged and Spiked, and Core II with Special Components were analyzed under various off-design conditions to induce nucleate boiling. The steady state code, STDY-3, written for the thermal analysis of pressurized water cores, was employed for the analysis. The code performs a complete steady state parallel channel thermal analysis for both nominal and hot channels. Thermal characteristics of individual elements were investigated while changing the parameters of primary pressure or inlet temperature to introduce the phenomenon of nucleate boiling in the the core. Reduction of system pressures to 1000, 800, and 600 psia and increasing core inlet temperatures to 465 and 500 degree F were studied as the means to induce boiling in the core. This analysis indicates that SM-1 Core I Rearranged and Spiked can be safely operated at the reduced pressure of 910 psia without introducing extensive boiling in the core. SM-1 Core II with Special Components can be operated at 800 psia or at an inlet temperature of 500 degree F at 1200 psia.

Abstract: A cyclic stress analysis of the SM-1 primary system was carried out. Problems encountered in the fabrication of PM-2A Core II and SM-lA Core II are described, and the results of an examination of damaged SM-lA Core I stationary fuel elements reported. A preliminary study of the radiation damage to SM-1 reactor vessel was made and the possibility of annealing the vessel discussed. Performance analyses are presented for five cores: SM-1 Core, SM-1 Core 1 rearranged and spiked, SM-1 Core II with special components, PM-2A Core 1, and SM- 1A Core 1. Preliminary critical experiments were made with SM-2 elements in a SM- 1 core configuration and nuclear and thermal analyses of the use of SM-2 elements in SM-1, SM-1A, and PM-2A completed. A throttling steam calorimeter was selected for measuring moisture carry-over on the PM-2A steam generator. Test procedures for evaluating the shielding of the SM-1, SM-lA, and PM-2A plants are summarized. Radiochemical and chemical analyses of SM-1 coolant and crud are summarized, and methods of activity control discussed. Preliminary results of studies of the properties of reactor pressure vessels under irradiation and no irradiation conditions are summarized briefly.