Recently a number of authors have suggested modifications of the Brueckner theory of nuclear matter so as to include hole-hole interactions, as well as particle-particle interactions. Iwamoto has demonstrated that in a perturbation theory calculation the inclusion of hole-hole interaction makes no change in the ground-state energy through second order. The singular two-body potential between nucleons makes it difficult, however, to conclude anything about the contribution of these terms in nuclear matter. The formal similarity between the equation of Iwamoto and the equation for the energy gap in nuclear matter, coupled with the fact that the energy gap is very small at normal density, indicates that the effect of hole-hole interactions is probably only a very small change in the ground-state energy of nuclear matter. It is the point of this note to show that this conclusion is in fact correct, the demonstration proceeding by use of the separation method for evaluating the energy of nuclear matter.

The solution photochemistry of chlorophyll and chlorophyll analogs is described. Many cases of electron transfer to or from the porphyrin macrocycle have been found, but in no case has any very large degree of energy storage been achieved. Because of the very rapid back-reaction for products with a {Delta}F of approximately -30 kcal, some solid state models in which such an energy storage might be achieved are described and their possible relation to the natural photosynthetic apparatus is given. We can see that while the solid state model (phthalocyanine) allows an approach from a somewhat different point of view, the net result is the same as what was sought, but so far not found, when we looked at the solution chemistry of chlorophyll (and chlorophyll model substances), namely, the transfer of an electron, or hydrogen atom, from the excited porphyrin to an electron acceptor at a high reduction level which can be used to reduce the ultimate carbon dioxide reducers, followed by the donation of an electron ultimately from water to the remaining radical ion, or lattice, which produces the net results of the transfer of the hydrogen from water to carbon dioxide.

This report discuss the design of the Plutonium Reclamation Facility the capacity of which will be 300 kilograms per month or 3600 kilograms per year or plutonium. The subject facility, as the name implies, must be extremely flexible in its ability to handle a wide variety of feed materials. The new facility will be operating on a three-shift day, five-day week, 40-week year with an overall efficiency of 75 percent; twelve weeks per year will be required for ``turnaround`` time to enable campaign operation for segregation of feed plutonium by isotopic content.

The purpose of this report is to define the extent of modifications necessary to increase capacities of the 100-B, C, D, DR, F, and H water plants for reactor flows of 90,000 95,000 105,000 and 115,000 GPM, and to provide supporting data for budget studies for increased production.

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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.

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.

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 …

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.

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 …

The objective of this test is to evaluate the hypothesis that autoclave film damage is a necessary condition for groove pitting of X-8001 alloy fuel cladding.

The purpose of this memorandum is to transmit revisions to the goal exposure plans for C-II-N, C-II-E, K-III-N, K-III-E, O-II-N and O-II-E material. Inherent in these plans is an assumption that the only restraint on metal usage is that imposed by the economics of the plutonium production process and associated uranium cycle. In the case when metal throughput is limited, by ex-reactor physical capabilities, to a level lower than the unrestrained reactor requirements, exposures higher than those recommended in this document would be indicated.

The objective of this test is to determine, through in-reactor testing, the resistance to corrosion of nickel plated fuel elements, plated by a chemical deposition technique. This program may eventually be composed of the following portions, the first of which is authorized by this test: (1) Irradiate ten columns of 0.5 mil chemically nickel-plated C-64 OIIN fuel elements having various heat treatments, each alternated with x-8001 control pieces. Two columns will be exposed to 400 MWD/T and eight to 800 NWD/T. (2) Should results from the above test be encouraging, separate authorization for the following tests may be requested: (a) Irradiate four columns of nickel-plated fuel and four of X-8001 clad fuel, both groups having purposely cocked pieces, until two ruptures are sustained in each group or until a factor of improvements of 400 is demonstrated at the 95% confidence level, unless the nickel-plated elements fail first. (b) Irradiate five columns of chemically nickel-plated C-64 clad OIIN fuel elements, alternating 0.2 mil and 0.5 mil plate, heat treated at 300{degrees}C for six hours. Two columns will be discharged at 400 MWD/T and three columns at 800 MWD/T exposure. (c) Irradiate approximately thirty columns of nickel-plated, C-64 alloy clad fuel elements …

The objective of this test was to irradiate self-supported fuel elements under conditions of severity comparable to those expected for future loadings of this geometry. From this irradiation an attempt was made to determine the resistance to corrosion of cooled cladding, the effect of supports on cladding corrosion and coolant flow patterns, resistance of ``hot-spot`` type of attack, and rupture resistance of projection fuel elements. These results are compared to rib supported control elements in a relative manner. To accomplish this comparison, four shortened columns of 1.47% enriched self-supported fuel were charged into B Reactor on 11-27-59, along with four columns of 1.47% enriched, current geometry fuel under authority of reference 1. As authorized, operating conditions were balanced as closely as possible within equipment and calculational limitations to minimize any ``built-in`` differences between the two materials being compared in order that exposure could be used as the basis for the rupture and corrosion comparisons.

It is the purpose of this document to provide assistance to the Manufacturing Section personnel in determining their future operating plans. In general, the inter-relationship of such engineering parameters as flow, reactor orificing, rupture performance, etc. has been considered. The effect of these engineering parameters are summed up in our recommendations for {open_quotes}Operating Plans{close_quotes} shown graphically in this document. It is to be emphasized that these plans do not reflect operational considerations which may modify the desirability of the indicated level increases nor has allowance been made for ability of the indicated level increases nor has allowance been made for major projects, major maintenance outages, or major changes in pile loadings. Many factor, which only Manufacturing personnel are capable of evaluating, may make it desirable to operate below or above these {open_quotes}Operating Plans.{close_quotes} These {open_quotes}plans{close_quotes} are based on incremental metal cost and burnout cost estimates obtained recently from L. W. Lang. A change in these assumed costs would require a revision to be made to these {open_quotes}plans.{close_quotes} It is also to be noted that many of the engineering parameters and basic assumptions which have been factored into these {open_quotes}plans{close_quotes} are subject to continual re-evaluation and revision. Thus, in a strict …

A design was developed for leading thermocouples from a high-temperature, pressurized water reactor-coolant system of such integrity that no reactor shutdowns were caused by its use. Using this design, measurements of the fuel-element-cladding temperature and its variation with time were made in three tests on elements clad in type X-8001 aluminum alloy. The following conclusions were reached from the test results: (1) the cladding temperature of a fuel element operated at low heat flux in high bulk-outlet temperature water did not increase with time and was slightly lower than predicted by the Sieder-Tate equation; (2) cladding temperatures of fuel elements operated at high heat flux in either high bulk-inlet or outlet temperature water increased 40 C higher than predicted by the Sieder-Tate equation with initial temperatures equal to the predicted temperatures; and (3) the rate of temperature increase appeared dependent only on fuel-element heat flux and location with respect to the front and rear faces of the reactor.

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The design approach followed in this study is that of handling and processing plutonium solution and powders within tightly sealed but readily replaceable units of processing equipment, each one of which will do one step in the overall process. The presently used method is to use open equipment enclosed in glove boxes attached to a transporting device. Elimination of the glove box enclose and its replacement by tightly sealed and detachable equipment provides the accessability necessary for equipment repair and reduction of down time, thereby reducing costs. Most of the primary functions for the entire button line have been studied and no insurmountable design problems have been encountered. Studies of the pre-reduction and filtrate treatment functions and the PR can unloading and RC can loading equipment have been summed up and compiled into a design criteria. Design concepts for all of the equipment for processing from the nitrate through the button stage have been worked out and appear to be acceptable approaches, except for the vacuum drum filter and the vibrating tube reactor. The latter two components have been studied and drawings prepared; however, the concepts developed are not considered entirely acceptable from the operation and maintenance standpoint. Because of …

In the past two years a number of process optimization studies have been made with the aid of IBM machine programs. These programs have been designed to calculate optimum power levels and optimum goal exposures so that better insight could be gained for use in resolving plant operating plans. We feel that this approach has been highly successful. In January of 1960, a new program was written for the IBM 709 to incorporate the desirable features of the previous programs and to provide additional output data. This document is a description of the new programs.

This report is an interpretation of data (flow, pressure, temperatures within a process tube during events affecting single tubes) as applied to the most severe (rapid) K-reactor transients which are credible. Analyses indicate that no fuel channel burnout will result from a BPA power loss to the process pumps.

The failure mechanics of high exposure, in-reactor coextruded fuel rods are quite different from those of defected unirradiated rods. The appearance and corrosion behavior of the high-exposure in-reactor failures suggests that the strength of the coextruded Zircaloy-2 to uranium bond has deteriorated. Notch-fracture tests, in which the strength of the Zircaloy-2 clad to uranium bond is evaluated in a qualitative manner, suggests that the bond strength has not deteriorated to the degree indicated by the failure behavior. It is believed that the irradiation induced property changes of the uranium fuel and not a deterioration of the character of the bond are responsible for the difference in irradiated and unirradiated failure behavior.

The data consists of out and inlet temperatures, and panellit and header pressures, sorted according to date and tube-pile.