The bremsstrahlung emitted by an electron scattered in a Coulomb field was first calculated by Bethe and Heitler. The total cross section for production of photons with wave number between k and k + dk by a nonrelativistic electron of kinetic energy {epsilon} is d{sigma}/dk dk = 16/3 Z{sup 2}r{sub 0}{sup 2} (e{sup 2}/hc) (mc{sup 2}/{epsilon})log ({radical} {epsilon}/hck + {radical} {epsilon}/hck -1) dk/k, where Ze is the charge of the (heavy) ion, and r{sub 0} is the classical electron radius. Bremsstrahlung in a plasma has been computed by a number of authors in the approximation of replacing the Coulomb field by a cut-off Coulomb or static Debye potential. It is the purpose of this communication to call attention to another important effect of the medium upon the rate of emission of bremsstrahlung. This may be described as a modification of the relation of the photon's energy to its wave number, due to the index of refraction of the medium. Equivalently, we note that one must include in the calculation of bremsstrahlung in a medium the photon-medium interactions which result in the 'clothing' of a 'bare' photon. The replacement of a particle by a quasiparticle has long been known to be …

None

This manual covers the technology involved in the 200 Area process for the recovery of Np/sup 237/ from certain aqueous waste streams in the separations plants, for the recovery of NP/sup 237/ and Pu/sup 238/ from irradiated NpO/sub 2/-Al slugs and for the fabrication of NpO/sub 2/-Al slugs. The manual contains sections on the fundamental chemistry, the primary recovery of Np by ion exchange, the decontamination of Np by ion exchange, the processing of NpO/sub 2/-Al targets, the separation and purification of Np/sup 237/ and Pu/sup 238/, the finishing of Np, the preparation of NpO/sub 2/, the disposal of spent resin, and the safety aspects of the handling of hydrazine. The section on the fabrication of NpO/sub 2/-Al slugs will be added later. 76 refs., 22 figs.

The purpose of this January, 1963 report is to review existing Redox exhaust gas sampling equipment and to propose modifications and additions where appropriate. A study was conducted to determine what modifications, if any, should be made to the Redox exhaust gas sampling equipment to improve their present methods of monitoring for excessive amounts of fission product activity.

Fourteen NAE1 fuel elements were discharged from KER Loop-4, after irradiation to an average exposure of 1250 MWD, at prototype N-Reactor coolant temperature and pressure. The elements were disassembled and measured in the KE fuel examination facility. This report includes all measurements, except the profilometer data.

Information is presented on current reactor fuel development (uranium core production data, fuel performance, AlSi process development, and alternate process development) and NPR fuel development (fuel production status, KER loop testing, and process development).

The information presented is a collection of most of the physics information available for the New Production Reactor. The details of some of the physics information, particularly those dealing with exposure and temperature effects, are by no means to be considered the final word since there has been no experimental verification of these effects. However, the gross physics characteristics described are felt to be reasonable representations of the expected physics behavior of N Reactor and should serve as useful guides throughout the startup planning and initial operation.

This invention discloses a method of making a synthetic nuclear fuel by mixing pure U-233 with U-238 fuels, such as depleted uranium, to produce a fuel that combines the most desirable nuclear properties of both the U-233 and U-238 fertile material. The process requires that U-233 be manufactured in such a manner as to be relatively uncontaminated from U-232. The subject of this invention is a method of fuel handling that incorporates the desirable features of both U-233 and U-238. It is proposed that U-233 be formed by short-term irradiations of thorium that will minimize contamination by U-232. At the end of a short thorium irradiation the U-233 is removed and combined with U-238. This combination then makes a synthetic fuel that has all the advantages of each ingredient but does not run into significant U-232 contaminant problems.

Production test IP-536-A authorizes the irradiation of N Reactor inner fuel tubes in KER-1 and KER-2. Supplement B, PT-IP-536-A, authorizes the irradiation of N Reactor inner fuel tubes prepared from a U-2 w/o Zr alloy. Both tests require that specific operating conditions and trip settings for particular loadings be approved by the managers of the Process and Reactor Development Sub-Section and the Process Technology Sub-Section prior to charging. The purpose of this document is to provide the process tube loading, high and low temperature operating limits, and water shutoff times for a charge of twelve 17.3-inch N inner fuel elements or twelve 18.5-inch U-2 w/o Zr elements in KER-1 and KER-2. The operating conditions were prepared for the 17.3-inch elements and with the same downstream spacer column, are conservative for the slightly longer U-2 w/o Zr charge.

This report details Hanford Production Reactors operating limits and performance and operating restrictions.

Part 1 of this two-part report presented a comparison of the irradiation-induced contraction rates obtained on standard-size graphite samples and the apparent moderator contraction rates of H, DR and C Reactors. The distortion trends of the top center of the graphite stacks, from which the apparent moderator contraction rates were derived, were presented as a function of cumulative power generation since start-up for each of the six older Hanford reactors. Since KE and KW Reactors are of identical construction, different in many details from the six older reactors, and have been operated under similar conditions, the data from KE and KW are treated as if from one reactor. As in Part 1, contraction rates apply to measurements made transverse to the extrusion axis of the bars. As used in this report, the depression rates refer to the rate of reduction in height of the top center of the reactor moderators. This depression rate is the net effect of contraction and expansion of the graphite bars. The depression rate is used to calculate the contraction rate of the individual bars. The contraction rates derived from the reactor data are compared with contraction rates of small samples of the same graphite (TSGBF) …

A study is in progress in which reactor operations will be simulated on the computer, the primary purpose being to evaluate the costs associated with various administrative alternatives which may be followed in conducting the overall operation, in addition to defining how operational costs are affected by such things as changes in fuel quality. The problem is complicated by the fact that the operation of the reactors is largely affected by random occurrences; no one can predict exactly when a tube will leak, or when a rupture will occur. Therefore, basic to the study is a probabilistic function, or set of functions, which govern the random aspects of reactor outages, and which can be used in the simulation study to generate reactor outages. This report is concerned with the derivation of such probabilistic functions. Although derived specifically for the simulation study, they are deemed of sufficient interest to warrant a separate report. Similar documents will be issued from time to time as the study progresses, and as results are found which are considered worthy of reporting prior to completion of the study.

This report summarizes and highlights the more importan milestones, development programs, performance and characteristics, and properties of dingot uranium and its use in the Hanford reactors from 1955 to the present. For the benefit of those unfamiliar with the terms ingot and dingot uranium as used in this report, ingot uranium refers to the metal made at the Fernald Plant of the National Lead Company by remelting under vacuum a mixed charge of solid scrap, briquetted scrap, and derby metal (the product of the UF{sub 4} bomb reduction), while dingot metal refers to the metal made by Mallinckrodt Chemical Works by reduction of UF{sub 4} to metal which is used directly (after suitable forming into convenient size for rolling), without the intermediate vacuum remelting step.

This report documents information acquired from the operation of the purex process during the year of 1963 at the Hanford Site. The report is broken down into sections containing data collected on two to four week intervals. The data included: Performance, flowsheets, feed preparation, solvent extraction, product treatment, solvent treatment, acid recovery, waste concentration, fission product recovery, and waste treatment and storage.

The metastability of alpha plutonium above the {alpha} {r_equilibrium} {beta} equilibrim transformation temperature (112 C) was studied by metallographic techniques. The purpose of this study was to extend the previous work on the {alpha} {yields} {beta} transformation which was accomplished using fluid displacement techniques{sup 1} and to obtain a basis for determining the length of time the alpha phase is stable at high temperatures. Approximate isothermal reaction curves were determined experimentally and an {alpha} {yields} {beta} time- temperature-transformation (T-T-T) curve was derived.

Plutonium-aluminum fuel elements, which were fabricated at Hanford Atomic Products Operation for a special Atomic Energy Commission program, but were rejected for failure to meet specifications, were returned to Hanford for recovery. These elements, although contaminated during acceptance checking have not been used in the proposed reactor tests. This special material will be stored until recovery is completed. This report provides the critical mass control specification for the 7.35 weight percent Pu-Al alloy fuel element storage gas cylinder storage AREA, 234-5 building.

The N-Reactor will be charged with tubular fuel elements consisting of two components. Both components, an outer tube with a concentric inner tube, have a uranium core, coextruded with a Zircaloy-2 cladding. The testing in support of N-Reactor fuels has been conducted in the experimental loops in the K-East Reactor. Although the process tubes in the KER loops are smaller than the proposed N-Reactor process tubes, the test loops operate at the water temperatures and pressures that will exist in the N-Reactor. To obtain the desired heat fluxes and core temperatures in KER loop test elements, enriched uranium fuel was used. The cladding surface temperatures and fuel temperatures were near anticipated N-Reactor values. Test conditions, results, and conclusions are presented.

The transformation of retained phases of plutonium of different quality has been determined over a period of 10--12 months, by observing the density variation with time. A large density increase of each specimen was observed. The density increase of the low quality metal was the greatest and that of the highest quality metal was the least. The magnitudes of these increases are apparent from the graphs. One of the specimens fractured into two pieces after ten months. Also, surface cracks appeared in two of the other specimens. The fractured surfaces probably originated from the stresses associated with the transformation of retained phases or possibly from stress corrosion. The density increase with time was the same for the as-cast and cold treated specimens.

During the early part of CY-1961, water leak frequency on plant began to pose a mounting problem. At this time, DR reactor exhibited a leak mechanism which, although observed to some extent at other reactors, appeared to be primarily a DR problem. This mechanism can best be described as a trench running longitudinally along the top portion of the process tube. The dimensions vary considerably; however, the lengths attained were generally about two to there feet and the width was roughly one inch. The beginning of the trench appeared to be mainly in the position of 11 to 14 feet fro the rear Van Stone flange. The radial position of the trench in the tube runs from about the 10: 30 to the 2:30 o`clock position. For lack of a more definitive expression for this corrosion, it will be referred to as a form of ``accelerated local corrosion.`` This report deals entirely with one possible means of affecting accelerated local corrosion. A layer of gaseous air coming from the dissolved air in the process water could form a heat transfer resistant area. This area would probably exhibit an accelerated corrosion rate due to the higher tube wall temperatures resulting.

A discrepancy of about 1000 gpm has existed between the full-flow recorded 190, 105 and ROL flows. While past operating practices have not used the 190 or ROL flow rates for official purposes, the disquieting, though not theoretically unexplicable, differences require some quantitative resolution. On November 24, 1962, a drawdown test of the 190-H storage tanks was performed to establish the accuracy of the various flowmeters. The drawdown test of the 190 storage tanks was run at the beginning of a scheduled reactor shutdown. With the full reactor flow supplied by the electric process pumps feeding from the storage tanks, the 183-H supply to the storage tanks was valved off. Additionally, non-process water usually taken from the storage tanks was valved off. The storage tank water levels were taken, then recorded as a function of time.

Three conditions representing various quantities of 10-inch and/or 12-inch fuel elements were presented by the Industrial Engineering Group, IPD, as a basis for initial estimates of economic incentive to partially convert the Production Fuels Section AlSi process to longer fuel elements for FY 1964. An earlier study (Ref. 1) pertained to 10-inch instead of 6-inch enriched fuels. The comparatively larger quantities of the 4 in. longer fuel for full E-N loading showed a larger payout than for the requirements shown in this study. Since the estimates have been prepared without the benefit any process development or equipment design, the validity of the results of this study rest directly with the assumptions. These assumptions are itemized in some detail to explain the basis from which payout was calculated. At best, the study can be used as a preliminary estimate, subject to changes as more detailed analysis are made.

The K Reactors were originally scoped to have three independent electrical power supplies and to be able to transfer process water, filtered water, raw water and electrical power from one K Plant to the other. It was required that shutdown flow to each reactor be available from at least four separate pumping stations. The originally scoped power level was 1300 MW, yet a plentitude of coolant supply means were provided, as can be seen from the above statements. Today, 1963, the K Reactors operate at 4400 MW and are capable of operating in the vicinity of 5000 MW during low inlet water temperature conditions. Requirements on coolant supply systems are more extensively defined today as a result of this larger potential liability and because of improved technology. This document discusses modifications to the K Reactors` secondary backup systems` equipment arrangement that will improve backup adequacy and reliability while permitting greater operating convenience as regards backup equipment maintenance. The modified secondary system proposed by K Maintenance Engineering is endorsed, provided a reliable means of avoiding overloading a boiler is included and provided the independence of the emergency generators can be reasonably assured. Preliminary discussions indicate that a control circuit to prevent …

This production test authorized the irradiation of 1.6% enriched single tube elements in KER-1 and 2. This document provides the process tube loading, operating limits, and water shutoff times for the charge.

This report details activities of the N-Reactor Department during the month of December 1962.