The objective of this irradiation is to further verify the corrosion rate of tubular-type fuel elements under conditions of high specific power and central core temperatures. This fuel will be the inner tube only of an NPR fuel assembly. As in previous tests, this inner tube rupture will be used to further substantiate the rupture detection instrumentation that is being used in the development of the NPR. Previously unirradiated fuel will be used in this test. The reactor is to operate at full power during the test. Permission is requested for charging two tubular elements The top element will have attached to it a hydraulic mechanism for opening a defect in the outer surface of the tube. The second or bottom element, will be used as a heater element to maintain loop temperature.

None

This report details activities of the Hanford Atomic Products Operation (HAPO) in 1959.

In Hanford reactors the reactivity gain upon loss of coolant water is an important factor in the speed of control requirements. The reactivity gain in the cold, clean reactor is determined from experiments, but additional effects must be taken into account if the gain in the operating reactors is to be obtained. One of these effects is the change in Xenon and Samarium poisoning with neutron temperature, which is discussed here. Earlier work on the relationship of operating limits to the reactivity gain upon loss of coolant is given in Reference 1. Work on this problem is continuing by Reactor Physics, IPD, but the newer work is not yet documented. In earlier calculations, the neutron temperature could only be guessed. Recent measurements of neutron temperatures have indicated the magnitude of the neutron temperature change upon loss of water. This document interprets the data of Reference 2 in terms of the change in Xenon and Samarium poison to be expected on water loss under typical operating conditions.

Object of this paper is to present preliminary results of experiments in Hanford in-reactor loops to determine if exposure to neutrons will increase corrosion rates of Al alloys, Zy-2, and 304 stainless steel. Results were negligible or no corrosion.

Irradiation of a third thermocouple train with a thermocouple element of the same design as used on the previous two trains, but with different heater elements, is authorized to an exposure no greater than 1000 MWD/T. A second decontamination of KER-1 using the same procedures as for the first decontamination is also authorized.

This document provides the loading and operating conditions for eight 16-inch or five 24-inch elements, either natural NIN-1 or .947% enriched NIE-1, in any of the four KER loops. The loadings are tabulated; operating conditions for either charge in KER 2, 3, 4, or 1 are given in figures.

Changes in operation and corresponding changes in the reactivity status of Hanford reactors are the result of a continuing effort to improve operating efficiency. Trends data related to these changes in operation and reactivity have been published previously for the periods from 1950 through 1958. The purpose of this report is to include trends data for 1959. Bar graphs in the first part of the report show yearly averages of selected data, and tables in the last part of the report show maximum, average, and minimum values. This document presents trends data for B, D, DR, and F reactors while a second document, HW-64932, presents trends data for C, H, KE, and KW reactors. Data included in past years which have not been included in this report are trends in pile power level at shutdown omitted due to a security status change regarding power levels, and number of temporary poison columns per startup omitted due to virtual elimination of temporary poison startups at B, D, DR, and F Reactors; added were potential non-equilibrium gains and potential equilibrium gains. Notice that all reactivity values are listed in the unit per cent excess k.

This document is the third of a series of question and answer lists issued as a review of the material discussed in the Reactor Processing Fundamentals Course. Each document represents the material covered by the Reactor Specialists during a typical three-month training program. Each question is discussed individually and the entire list completed during the three-month session. Each three-month-training period is devoted to the complete discussion of a single chapter of the IPD Physics Primer series. The questions are compiled in a logical sequence with the material as presented in Chapter V of the Primer series control. As the course progresses, a certain amount of recall of earlier chapters is essential to a thorough knowledge of the specifics of reactivity and distribution control. Therefore the first half of this document consists of material previously presented in a slightly different manner. Basic points are stressed; the intent is, of course, that informative discussions lead to a better understanding of the material presented in the Primer.

This report documents analysis of radionuclides from reactor effluent water samples. Separation from the sample matrix is accomplished by successive carbonate, hydroxide and fluoride precipitations. Experimental data is provided.

This report discusses the use of liquid ammonia as a tracer for the detection of cooling water leaks into the piles. It is both safe and cheap and can be detected by methods adaptable to direct-reading instrumentation on a continuous-flow sample. Moderate capital costs and materials costs of less than $50 per pile test are anticipated.

The assumption of tube replacement losses can have a significant effect in the determination of optimum power levels and goal exposures. The tube replacement loss term in the reactor process optimization model is based on calculating the average projected tube replacement losses that will result from operation at given power and temperature conditions. Tube replacement losses associated with external corrosion, Van Stone flange failure, etc. (i.e., basically non-power level-temperature dependent) are assumed to be miscellaneous losses and are not included in the tube replacement term. Over a period of several years the experienced tube replacement losses (due to internal corrosion) should compare reasonably well with the losses predicted by the optimization model. Planned tube replacement project action which would require, in some cases, premature tube removal is a special situation which requires a modified approach to the prediction of tube replacement losses and to the optimization of reactor power levels prior to project action. A method has been developed for optimizing reactor power levels and goal exposures based on any assumed calendar date for major tube replacement project action. It is the purpose of this document to discuss the general application of this method in the optimization and illustrate the …

The purpose of this report is to establish the scope of a project to provide a system which will quickly and accurately identify the sources of all reactor scrams.

This report investigates a study of the accuracy of the power level calculations. Some characteristics of this calculation can be adduced but the accuracy in the sense of freedom from bias leads to an impasse. It presumes that there is another measurement for comparison purposes for which there are engineering reasons to believe, the inherent error is smaller. For the power level values as calculated by the Foxboro Power Level Calculator no such measurement or calculation exits. However, the calculations can be studied to determine whether it is worth-while to make further refinements relative to the various functions of the power level calculation. The accuracy requirements vary according to use where the most stringent requirements are in the control of SS Materials.

The 190-KW process water pumping Unit No. 1 is scheduled to be modified for increased pumping capacity under Project CGI-883- Component parts for this modification are expected to-be received during June 1960. Installation of these components would require approximately ten days; due mainly to grinding of the high lift pump case to make room for the new larger diameter impeller. In order to minimize lost production, it has been proposed by K Reactor Operation that the high lift pump be modified early this spring during the scheduled maintenance overhaul period on pumping unit No. 1. The test impeller recently removed from the No. 1 high lift pump in KE would be repaired and installed in the KW pump at this time. Later, in June or July when the components for the complete modification are available the low-lift pump and drive motor would be modified and associated electrical and instrument alterations would be completed during a normal reactor outage. Adoption of the proposed plan would make it necessary to operate the modified high lift pump for a period of approximately three months with an unmodified low-lift pump. A study was made to determine the feasibility of operating the pumping unit in …

The high pressure loop installed in the 3X3 reflector position of the ETR and the associated instrumentation to detect and study failure mechanisms handled the rupture tests without difficulty. Failure of the elements was initiated by shearing off a projection on the fuel elements. The first test of the series used previously unirradiated seven-rod clusters. After the projection was sheared off the fuel elements were operated for seven hours with no failure. Failure is defined as having occurred when sufficient uranium oxide has formed to split open the cladding and release large amounts of fission products into the loop water. The second and third tests used fuel which had been irradiated to 2400 MWD/T at Hanford prior to insertion into the ETR. The second test was operated for 14 hours after the projection was sheared off--again with no failure. The third test was operated for only 33 minutes after the projection was sheared off before fission product activity in the loop water caused the test to be terminated.

Cavitational flow has been known to exist in-rear crossheader ``Parker`` fittings at B, D, and F Reactors for the last five or six years. Calculations showing initiation of cavitational flow as a result of high flow rates in the present fittings were verified by experimental data in 1954. A study is currently being conducted to determine the required plant modifications to obtain flow increases on the order of fifty percent above existing flows. This study and the results of preliminary tests that show nominal flow increases may be obtained by reaming rear crossheader fittings has focused attention on the condition of existing rear face piping. To obtain an estimate of the effect of cavitation in rear crossheader fittings resulting from past and current operating conditions, twenty one fittings were examined during the period October 6, 1959 to November 30, 1959. This document reports the inspection results.

G. A. Last reports on metallic uranium irradiation studies: cluster irradiations, tubular element irradiations, MTR-ETR irradiations on NaK capsules, effects of pH on film formation, fuel performance, effect of pressure on swelling, and ETR rupture tests. R. D. Leggett reports on irradiation program, and pore site and distribution in irradiated uranium.

Cobalt samples frequently are irradiated in nuclear reactors to produce gamma sources and can be irradiated as integral flux monitors because of the long half-life of the isotope produced. At the present time a small cobalt sample is being irradiated within the KW Reactor Snout facility for future use as a radiographic source for inspection of finished product in the Chemical Processing Department. Analysis was made to estimate the buildup of activity in this sample; the general equation may be of interest and value for other cobalt sample irradiations.

The possibility was reviewed that errors in calculations might have resulted in an erroneously high theoretical strontium activity which would explain the unexpected low activity found ion the strontium recovered from Purex wastes. The new, corrected calculations (49,213 curies/T) improves the accounting from 65 to 75% of theoretical. The discrepancy still should be investigated.

The objective of this test authorization is to increase reactivity and thus reduce short term enrichment requirements by increasing graphite temperature limits during low exposure operation following full central zone discharge.

Since startup in January 1945 through December 1959, 4 {times} 10{sup 9} gallons of radioactive liquid wastes have been discharged to cribs and trenches at HAPO by the Chemical Processing Department facilities in the 200 Areas. These wastes contained approximately 2.5 {times} 10{sup 6} curies of beta emitters. The Scavenged Waste Recovery Program was completed in 1957 and Redox Plant process changes were made during the latter part of 1958. These changes resulted in significant reduction in the amount of radioactive materials that have been discharged to the ground in subsequent years, from a maximum of 8.2 {times} 10{sup 5} curies in 1955 to 9 {times} 10{sup 3} curies in 1959. Although these large amounts of radioactive materials have been discharged to the ground, periodic waste report inventories include no reduction due to radioactive decay. The estimated depletion by radioactive decay is the basis of this report.

The projected E-N load at H Reactor will require a complete change in the type of charge placed in most flattened zone process tubes. Objective of this test is to determine the reactivity of the proposed E-N charge relative to the known natural uranium charge reactivity.

In-reactor tests of self supported fuel elements is ribless process tubes have shown a significant reduction in hot spot failure incidents. It is believed that the support rails prevent gross fuel element cocking or misalignment which would allow inadequate cooling of a portion of the fuel element. It is reasoned that if similar spacing devices were attached to fuel elements for use in present ribbed tubes an appreciable reduction in hot spot failures would result. The problem then remains to select suitable spacing devices hereafter called called ``bumpers`` and to assess the increased energy losses associated with their use. Efforts to predict the energy losses which may be caused by fuel element support rails on bumpers have been somewhat discouraging. The odd shape of the support rails which appear something like a close-coupled suitcase handle preclude the rigorous use of available drag coefficients or contraction expansion coefficients. Hence it is necessary to make pressure drop measurements with a fuel elements design (size) which is considered to be quite close to meeting the actual pressure flow requirements. Then the final design is determined in view of these data. The data of this report were obtained in the 189-D Hydraulics Laboratory to …