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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.

Americium (atomic weight 241) grows into plutonium as a result of the 13-year half-life beta decay of Pu{sup 241}. An appreciable quantity of Am{sup 241} has.grown into the scrap that accumulated following shut-down of the Recuplex facility, and which will be processed during the initial period of operation of the plutonium reclamation facility. Current interests in trans-plutonium research and in isotopic heat sources make it desirable to consider the recovery of this Am{sup 241}. The americium contained in scrap that is processed through the reclamation facility should appear almost quantitatively in the aqueous waste stream (CAW). Subsequent processing of the CAW through the DBBP extraction column (CW) of the waste treatment facility should result in further separation of plutonium and americium by extracting some of the plutonium while most of the americium remains in the aqueous waste stream (CWW). The fact that a low free nitric acid concentration will favor the extraction of americium into TBP-type solvents can be used to recover americium from the large volume, high salt concentration CWW stream. The flowsheets in this document present the chemical conditions for effecting this recovery.

Reactor water leaks cause outage time for locating and eliminating the leaks and for repairing damage paused by the vater that enters the reactor. The justification for equipment and process changes necessary for reducing the number of leaks and the resultant outage time is dependent on the number of leaks expected. This document presents an estimate of the future reactor water leak frequency and the average outage time charged to each leak. A water leak is defined as any entry of vater into the reactor moderator during operation. Future vater leaks are expected to result primarily from Van Stone flange failures and miscellaneous causes such as mechanical damage, fuel ruptures, and transverse cracks. It is assumed that continuous emphasis will be placed on corrosion monitoring to determine the tubes that must be replaced to prevent leaks because of internal and external corrosion. The number of water leaks is expected to average between 80 and 100 per year in the future. The normal-range of the number of leaks probably will be between 60 and 120 per year as a result of normal variation in the process tube physical condition. The number of outage hours charged per leak is expected to continue …

Test K-4-14 described in this report, was the second ``full charge`` test to be irradiated in KER-4 following loop modification under Project CGI-839. It provided an opportunity to accumulate additional data on the behavior of N Reactor fuel elements at conditions equivalent to or more severe than those anticipated in N Reactor during Phase I, equilibrium operation. Objectives and test results are provided.

The research and development activities reported on are: enriched fuel (1.25% U{sup 235}), flex 2 code development, Pu burning and recycle, single tube fuel element, heavy isotope production, U{sup 233} -- Th{sup 232} fuel cycle, physics input for economic studies, Pu{sup 240} effective resonance integral, N-Reactor operator certification lectures, bases for control and nuclear instrumentation, and shield evaluation. (GHH)

The purpose of this supplement is to authorize charging the complete central zone core of H Reactor with the modified E-N charge makeup, installed as a 102-tube ``test core`` in the fifth loading. The potential economic and efficiency gains of a simplified charge pattern (a smaller average number of N pieces per tube and a minimum number of different charge types) were discussed in Supplement II to the parent PITA. Accordingly, a test block, consisting of 102 process channels in the center of the reactor, was included in the charging of the present (fifth) E-N core loading for the purpose of evaluation. Based on satisfactory preliminary confirmations of the calculated reactivity, front-to-rear flux distribution and other operating characteristics, extension to a full-scale core loading appears feasible. All resulting changes in pile reactivity and other operating variables will be minor and easily adjustable without significantly affecting operating efficiencies.

One of the problems which may be encountered in the operation of NPR, or of any similar pressurized-water reactor, is the deposition of crud in the reactor core, particularly on the surfaces of the fuel elements. This crud consists primarily of iron oxides resulting from the corrosion of the piping. It has generally been found to deposit preferentially in a radiation field such as is present in the reactor core. Such deposition is undesirable for two reasons. First, the crud will become activated, and on subsequent release may create radioactivity levels in the ex-reactor components of the primary loop high enough to be a hazard to maintenance and operating personnel. Secondly, the crud deposit is a poor conductor of heat, and even a rather thin film can raise the cladding temperature by 100 -- 200{degree}C and result in accelerated corrosion of the cladding.

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This document is submitted in response to a request from the Atomic Energy Commission for a report on the safety aspects of reactor operation at bulk outlet temperatures in the neighborhood of 95 C. To properly describe the safety aspects of operation it is necessary to address the report to the engineering parameters associated with reactor operation and compare them to actual technical limits. Also, some discussion of administrative power level limits and bulk outlet temperature is in order.

This report addresses the semiannual progress report for the period ending on February 15, 1963.

The information obtained in this experiment was deduced by measuring the pumping speed as a function of the energy of the incident electrons and as a function of the energy which ions hit the surface.