Abstract: An analysis of SM-1 Core II and SM-1A Core I zero power experiments was made by comparing these cores to each other and to AM-1 Core I on the basis of critical bank positions, bank calibrations and available chemical analyses of the fuel plate compositions. The effects of replacing boron absorbers by europium absorbers upon rod worth and stuck rod conditions were studied. Comparisons of measured and calculated power distributions were made. It was concluded that both SM-1 Core II and SM-1A Core I contain nearly identical B-10 loading of 17.79 grams, compared to the best estimate of 15.75 grams for SM-1 Core I. The available data indicates that all three cores possess similar nuclear characteristics.

Abstract: This technical report contains a description of the nuclear analysis performed upon the SM-2 core and vessel for the period September 1, 1958 to December 31, 1959. Calculations are given for core reactivity, power distributions, lifetime, reactor control, kinetics, radiation problems, fuel and poison burn-ups, and the nuclear effects of poisons, temperature, and geometry. Wherever possible, experimental data is included in order to test the validity of the analytical models. The SM-2 nuclear analysis was performed by Alco Products, in. under Tasks 1, 8, and 10 of Contract No. AT(30-2)-326 for the Atomic Energy Commission.

This report describes the work carried out in 321 Building semiworks equipment, to define the factors contributing to high corrosion-product contamination and presents recommendations for reducing the impurity level to meet current specifications (maximum of 200 parts total metals per million parts U).

Abstract: This technical report is concerned with the program and results of the SM-1 reactor vessel head removal and core inspection at Fort Belvoir, Virginia from the period March 7, 1969 to May 17, 1959. It covers the operating procedures in detail and records the conditions found and problems encountered in order to make a record for reference in future work of this nature. The major objective of the program, to obtain irradiation data on the SM-1 type core, has been met. The boron control rod elements were found unsatisfactory for full core life, and europium oxide elements were placed in the core for future irradiation stability data. A major problem was experienced with the cracked pressure vessel head studs. The methods developed for the removal of the broken studs are presented. The complete metallurgical study of the stud failure from stress corrosion is included as Appendix B.

Abstract: The results of activity buildup studies in the SM-1 performed during Core I lifetime (June 3, 1957 to April 28, 1960) are reported. Data are presented on the extent, nature, and mechanism of the buildup of long-lived gamma emitting nuclides in the reactor primary system. Radiation levels after reactor shutdown are presented, as well as mathematical equations used to account for the observed activity levels. The data have shown that Co60 is the major contributor to radiation levels in the SM-1. Co60 activity arises from the cobalt in Haynes 25 alloy flux suppressors, and the cobalt impurity in stainless steel. After 35 months operation at an average power level of 55%, deposited Co60 activity accounted for approximately 83% of the total radiation level (mr/hr) contributed by the long-lived gamma emitting nuclides. The contribution of the primary coolant activity to the total radiation level is insignificant when compared to the contribution of the activity deposited on the walls of the system. The radiation level on the super-heater side of the steam generator was about 1400 mr/hr after 35 months of reactor operation. The percentages of Co60 activity in the coolant and in the deposits were not the same. This indicates …

Abstract: A zero power experiment on the SM-1 Core II included an element by element reactivity check of fuel elements and control rod absorber sections, and an estimate of burnable nuclear poison loading in stationary fuel elements. An approach to criticality was made by the inverse multiplication method, and critical rod bank position obtained as a function fuel loading up to full core loading. Minimum and maximum core reactivity measurements were obtained by selective loading of stationary fuel elements, and total excess K of the core was established. Power distribution measurements were taken in the regions of the core-reflector interface and the fuel-absorber interface in the control rod assemblies. The effectiveness of europium flux suppressors in the top of control rod fuel elements was determined, and power peaking was measured in stationary elements adjacent to control rod assembly water gaps. Survey measurements established the work of spiking clean, cold SM-1 cores with SM-2 elements and the work of water holes in the SM-1 Core. The reduced scope zero power experiment performed on SM-1A core I included an element by element uniformity check of stationary fuel elements, a core assembly test, comparison of Eu2O3 and B4C absorber sections, and development of …

Abstract: Development of welding processes for stainless steel fuel elements, previously initiated by Alco Products, was continued in Task 5. Three objectives were accomplished; (1) development of a suitable reference welding process, (2) evaluation by out-of-pile tests of the structural integrity of reference welded fuel elements, and (3) preparation of specifications and fabrication procedures for the reference welding process.

Abstract: This technical report summarizes the data obtained in a recent literature survey conducted to determine the effects of neutron irradiation on the impact and other mechanical properties of both ferritic steels and austenitic stainless steels. The survey was primarily aimed at obtaining sufficient data on the behavior of pressure vessel steels at high integrated neutron flux levels in order that a reference material of construction could be selected for the SM-2 (APPR-1B) reactor vessel. Materials studied in this literature survey included carbon and low alloy steels such as: ASTM A-212B, ASTM A-201, ASTM A-301B (CR-Mo), ASTM A-106 (coarse and fine grained), ASTM A-285, ASTM A-302B (Mn-Mo), ASTM A-353, ASTM A-203 Grade D, E-7016 carbon steel weld metal, USS Carilloy T-1, HY-65 and HY-80. In addition, Types 304 and 347 stainless steels were also investigated as representative austenitic materials which might be used in pressure vessel construction. A careful evaluation was made of the irradiation induced changes in the mechanical properties of the above materials. The ferritic steels were evaluated primarily on the basis of increases in transition temperature due to irradiation and decreases in the amount of maximum energy absorbed prior to ductile failure. Factors such as industrial experience, …

Abstract: An element by element reactivity check for SM-1 Core II fuel elements and control rod absorber sections was performed and the burnable nuclear poison loading in the SM-1 Core II stationary fuel elements was established. An approach to criticality of the SM-1 Core II was performed by the inverse multiplication method and the critical rod bank position obtained as a function of fuel loading up to the full SM-1 Core II loading. Maximum and minimum core reactivity measurements were obtained by selective loading of stationary fuel elements and the total "excess K" for the core established. Power distribution measurements in the region of the core-reflector interface and the fuel-absorber interface in the control rod assemblies were performed. The effectiveness of europium flux suppressors in the top of control rod fuel elements and the power peaking in stationary elements adjacent to water gaps in control rod assemblies were measured. Survey measurements established the worth of spiking cold clean SM-1 cores with SM-2 elements, and of water holes in the SM-1 core which might be utilized as flux traps for materials irradiation.