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Scram transient tests PT-IP-249-C

Description: The purpose of this production test is to provide a standard method of obtaining scram transient reactivity information at the eight reactors, under conditions conducive to valid data. These conditions include the bypassing of the Panellit system at a low power level for a short, controlled period of time during May 1959.
Date: May 25, 1959
Creator: Bowers, C.E.
Partner: UNT Libraries Government Documents Department

Proposal for elimination of 100 percent inspection for uranium grain size

Description: Uranium cores have a grain size specification of A8 to A3. Large grain size is denoted by a small ``A`` number and small grain by larger ``A`` numbers. Grain sizes smaller than A9 denote an untransformed core which presumably has a high rupture potential. Grain sizes larger than A3 cause swelling, or bumping, of the fuel thus increasing the probability of rupture failure. Presently, tests are made on all cores received to determine compliance to the above specifications. As a result of a recent study and analysis of all pertinent factors applicable to this situation, it is concluded that 100% inspection of all lots received at HAPO for compliance to grain size specifications is neither warranted nor economically justifiable. In lieu of the present screening, a system is proposed to insure that screening is effected only when there is an economic incentive. Such a system requires that the Feedsite (NLO) maintain its present process control sampling plan of 5 cores per ingot. A sampling receiving inspection at HAPO will be used to detect major shifts in the quality of cores to insure that screening is used only when there is an economic incentive. This sampling inspection, in place of the present 100% screening, will result in an estimated annual savings, to IPD, of $52,500, based on the 0.0057% defective rate currently observed in the incoming material received at HAPO.
Date: March 30, 1965
Creator: Stevenson, W.F.
Partner: UNT Libraries Government Documents Department

Fuel element design handbook

Description: The economic development of nuclear reactors depends upon the integrated progress in the fields of reactor design, fuel element design, reactor operation, and fuel production and separation. Broad criteria, which restrict the fuel element design, are determined by the mutual consideration of the problems encountered in all the above fields. Hence, no stage of reactor design or operation is independent of the fuel element problem, nor can the fuel element designer disregard the interest of any one field. As an introduction to the fuel element design problem, this chapter describes how the general criteria for a fuel element are determined.
Date: September 1, 1958
Creator: Merckx, K.R.
Partner: UNT Libraries Government Documents Department

Technical activities report research and development & 234-5 metallurgy groups metallurgy - pile technology unit, October 1951

Description: This is a technical progress report for the Research and Development, and 234-5 Metallurgy Groups of the metallurgy-pile technology unit for the period Oct 1951. Numerous reports are attached under the general areas: P-10 alloy development; uranium metallurgy; metallurgy of Hanford structural materials; radiometallurgy, facility development; plutonium metallurgy program; canning development; plant service work; 234-5 metallurgy group, plutonium metallurgy program.
Date: November 1, 1951
Creator: Schalliol, W.L. & Wick, O.J.
Partner: UNT Libraries Government Documents Department

PITA-31 fringe-blanket irradiation of thorium oxide. Supplement VI

Description: The objective of this supplement is to authorize recharging of the thoria in the fringe zones of the B, C, D, KE, and KW Reactors. Initial charging is described in the parent document. This authorization will serve as an interim measure until Reactor Process Standards have been approved to authorize fringe blanket thoria irradiations.
Date: July 1, 1965
Creator: Gross, P.D. & Hladek, K.L.
Partner: UNT Libraries Government Documents Department

Numerical results of PT-IP-338A and supplement A: DR reactor heat decay test at high outlet water temperatures

Description: This report summarizes the results obtained in the experimental measurement of the reactor heat extracted by the coolant during two different scrams of the DR Reactor. These tests were conducted with the reactor flow carefully controlled to establish different reactor coolant temperatures following the scram. During normal reactor scrams, which have been used in the past to measure the reactor heat output, a high coolant flow was maintained. Consequently, the coolant outlet temperature dropped very quickly from an operating value of 90 to 95 C down to 20 to 30 C following the scram. Under emergency conditions, however, with only the last ditch emergency flow available, the coolant outlet temperature could remain as high as 80 to 90 C. This condition could prevail because of the small amount of last ditch flow available, and could last for 5 to as long as 70 minutes after the reactor scram.
Date: December 3, 1962
Creator: Jones, S.S. & Manuel, J.L.
Partner: UNT Libraries Government Documents Department

Design scope of the Z Plant Metal Control Facility, Project CGC-944

Description: The steadily increasing plutonium production rates in the 234-5 Building has resulted in a large increase in the number of PR cans that are now being handled in the building. This has resulted in the available receiving and storage places for PR cans being taxed beyond their full capacity thereby necessitating the storage of PR cans within the building corridors. Accordingly, the ten-year business plan for the 234-5 Building, as documented in HW-65000, included a program for providing new receiving and storage facilities to alleviate this problem. The purpose of this document is to present a complete process engineering design on the combined receiving, storing, and blending facility. This document will provide the basis for the preparation of all subsequent Title 1 and Title 2 designs of this combined facility which has been given the title of Metal Control Facility.
Date: March 12, 1962
Creator: Haberman, H.D.
Partner: UNT Libraries Government Documents Department

Elimination of TOA corrosion limits

Description: In 1958, planned large scale use of the new I & E slug geometry at more severe operating conditions than had been generally experienced suggested a possible compromise in reactor life and safety if a reasonable degree of rupture control with the new type of element was not maintained. The formalized slug corrosion limit (Top-of-Annulus limit) was issued as a Process Standard at the time of the full-scale loading of I & E geometry fuel elements to provide this limit for reactor operation. The loading of I & E slugs at all reactors has been accomplished and initial power level increases have been made. To date, 67 I & E ruptures have been sustained including both `hole` and `annulus` failures. The type and behavior of ruptures to be expected with I & E geometry are now characterized. Recent studies have indicated that the I & E failure experience is consistent with the general mathematical rupture model formulated from analysis of solid slug experience. Increased confidence in the use of this model in combination with Optimization Studies permits greater emphasis to be placed on the rupture model as a guide for reactor operation. It is the purpose of this report to present the basis for substituting the rupture model for the TOA corrosion limits for rupture control purposes.
Date: May 12, 1959
Creator: Graves, S.M.
Partner: UNT Libraries Government Documents Department

Examination of four irradiated KVMS fuel elements - RM C-440

Description: Four KVNS production fuel elements which had operated for 66 days were submitted to the Radiometallurgy Laboratory for examination after being subjected to unusual conditions during shutdown. It was believed that the elements may have been at elevated temperatures for as long as 4-1/2 hours after shutdown. An abnormal film which appeared to be boiler scale was observed on some of the elements when they were examined in the basin immediately after discharge. Four elements were selected for metallurgical examination to determine whether they had been overheated and to assess any resulting damage or change.
Date: September 1, 1964
Creator: Gruber, W.J.
Partner: UNT Libraries Government Documents Department

Graphite burnout, interim report on IP-25-A (PT-105-532-E)

Description: Graphite reacts with such gases as CO{sub 2}, O{sub 2}, or water vapor to form gaseous oxides of carbon. In the case of CO{sub 2}-graphite interaction, the reaction rate is not significant until about 550 C. Water oxidizes graphite, very roughly, three times faster than CO{sub 2}. Air will oxidize graphite appreciably at temperatures below 500 C. Graphite removal from Hanford reactors is very important, since graphite is used both as a structural support and a moderator for neutrons. Griggs has shown that small graphite samples oxidized to 10 per cent weight loss had only about one-half their original compression strength. Hence, the longevity of the reactors depends to a great extent on maintaining a low graphite oxidation rate. A means of monitoring the extent of graphite loss, i. e., the burnout rate, is necessary to establish future reactor operational standards. Presently, weighed samples of reactor grade graphite are placed along the length of an empty process channel in each reactor. Thus, a sample is exposed to the reactor`s ambient conditions of power level, moderator temperature, and gas composition. This program was initiated in the vicinity of June, 1953 by Woodley. This report presents data on graphite burnout obtained from in-reactor experiments authorized under IP-25-A (PT-105-532-E) from August, 1957 to January, 1960. Burnout rates are obtained by a direct measurement of the weight loss of control graphite samples exposed to the reactor atmosphere.
Date: March 15, 1960
Creator: Ryan, B.A. & Halas, D.R. de
Partner: UNT Libraries Government Documents Department

Special yield data analyses for core analysis material key {number_sign} 6887-KW-E

Description: The report consists of three tables listing the chemical and isotopic composition of samples named KW-C-2, KW-C-3, and pig. Plutonium 238, 239, 240, 241, and 242, and uranium 234, 235, 236, and 238 are given in weight percent. Samples vary from 1.8--3.2 lbs U/gal and 0.5--0.85 gms Pu/gal. Concentrations of Am-241, Cm-242, and Np-237 are also reported.
Date: June 16, 1965
Creator: Braymen, W.H. & Smith, H.E.
Partner: UNT Libraries Government Documents Department

Quarterly progress report, Metallurgy Research sub-section, April 1955--June 1955

Description: One uranium tensile specimen irradiated to 620 MWD/T (1.75 x 10{sup 20} nvt) was tested at 285 C (545 F). The values obtained were: ultimate tensile strength 71,000 psi, 0.1% offset yield strength 70,000 psi, percent elongation (one inch gage length) 0.7, and modulus of elasticity 12 x 10{sup 6}. These values are comparable to the as-irradiated values of specimens tested at room temperature which were: ultimate tensile strength 76,000 psi, 0.1% offset yield strength 71,500 psi, and percent elongation (one inch gage length) 0.36. The elongation of the specimen tested at 285 C was less than that of a specimen annealed at 700 C (1290 F) after irradiation, then tested at room temperature. Experiments have been initiated to determine the structural stability of irradiated uranium-silicon alloy when this alloy is in the stable epsilon phase. Present work is limited to the evaluation of thermal expansion as a criterion for degree of epsilonization. This technique has not proven satisfactory. The damage resulting from the bombardment of uranium with electrons in the 1-2 MEV range is to be evaluated using electrical resistivity and x-ray diffraction to determine the extent of the damage. Calculations indicate a threshold energy of about 1.25 MEV electrons is necessary to generate vacancy-interstitial pairs.
Date: August 15, 1955
Creator: Cadwell, J.J.
Partner: UNT Libraries Government Documents Department

KER loop fuel testing program and schedule, CY 1962

Description: The interests of several departments at Hanford are involved in the planning, execution and evaluation of the results of the KER loop testing effort in support of the NPR fuel program. The varied interests and activities of the participating groups must be well-integrated if effective use of our limited testing capability is to be made. The purpose of this report is to help achieve this integration by summarizing the current thinking on the goals of the NPR fuel testing program and by presenting the current loop schedule.
Date: February 14, 1962
Creator: Evans, T.W. & Kratzer, W.K.
Partner: UNT Libraries Government Documents Department

Technological hazards evaluation core E-Q load

Description: The United States Atomic Energy Commission authorized the production of 110 kg of U-233 by the irradiation of thorium oxide in the Hanford production reactors. This irradiation will be carried out by replacing the natural uranium fuel in all or a portion of the process tubes in several of the reactors with thorium oxide (thoria) and slightly enriched uranium. This reactor loading is referred to as an ``E-Q load.`` Reactor loadings are separated into two types which can be charged independently of each other: a mixed lattice in the central zone of the reactor in which enriched uranium (0.947 per cent uranium-235) and thoria are loaded in separate process tubes in uniform array with a fuel-to-target ratio of about 5.5--6.0 to 1 (core E-Q load), and a peripheral ring of thoria in the outermost process tubes of the reactor with the reactivity being supplied by enriched uranium (0.947 per cent uranium-235) in the process tubes in the immediately adjacent two lattice units (blanket E-Q load). This safety analysis is addressed to the core E-Q load; however, the specific fringe loading, with properly matched reactivity, has little effect on the basic physics and hydraulics of the reactor so the evaluation presented here is applicable to a combined blanket and core E-Q load. The evaluation applies to both the K and smaller production reactors.
Date: October 6, 1964
Creator: Greager, O.H.
Partner: UNT Libraries Government Documents Department

Production test 231-12: Recovery of americium. Final report

Description: In response to the need for isolating small quantities of americium for employment in research studies, principally at laboratories not connected with the Hanford Atomic Products Operation, a program was formulated to accomplish this at minimum cost and disturbance to plutonium production. This report describes the operations employed and the results obtained in the isolation of americium from aged plutonium nitrate solution utilizing the Isolation Building process equipment.
Date: March 2, 1954
Creator: Packer, G.V.
Partner: UNT Libraries Government Documents Department

Quarterly progress report - metallurgy unit - April 1954--June 1954

Description: Four irradiated tensile specimens were tested in the Radiometallurgy laboratory. The uranium was from a lot rolled at Simonds Saw and STeel Company, then beta heat treated by the triple dip process. The samples were irradiated to 310 MWD/AT. The corrected exposure was 620 MWD/T after compensating for the decrease in flux depression due to the specimen size. Two of the irradiated specimens were tested in the as-received condition while the other two were vacuum annealed at 400{degree}C for 15 hours prior to testing. When the tensile properties of the unirradiated controls, irradiated, and irradiated-annealed specimens were compared, it was observed that no marked change in modulus of elasticity occurred, the 0.1 percent offset yield strength increased by a factor of about two for the irradiated and irradiated-irradiated, and 35 for the irradiated-annealed specimens, compared to the controls. The marked change in mechanical properties on irradiation as well as the relatively small effect of the anneal, indicates that the foreign atom effect caused by fission product concentration and distribution is the controlling factor, and work-hardening type of damage is relatively insignificant.
Date: July 16, 1954
Creator: Cadwell, J.J.
Partner: UNT Libraries Government Documents Department

Authorization for a process change -- Revision of nuclear safety rules for transporting plutonium compounds between the 231 and 234-5 Buildings

Description: Nuclear safety rules for transporting plutonium compounds between the 231 and 234-5 Buildings were established in Documents HW-30328 and HW-31465. Provisions for the co-transporting of loaded sample cans, filter boats and/or product samples were not included in the above documents. In the interest of economy and efficiency the rules have been reviewed and revised according to the provisions set forth in this document. The paper describes the present transportation limits and the revised limits for transporting plutonium nitrate, plutonium oxalate, and other plutonium samples. The limits are in keeping with the policy of preventing the accidental accumulation of unsafe masses of Pu.
Date: June 9, 1954
Creator: Smith, A. E.
Partner: UNT Libraries Government Documents Department

Hypothesis concerning irradiation embrittlement of uranium

Description: In discussions with a number of people, a hypothesis has been evolved which appears to fit available information concerning irradiation embrittlement of uranium as well as indicate a possible solution to the problem. The purpose of this memorandum is to expound this hypothesis as an aid to those working with the problem. Since it embodies the ideas of many people, no claim to unique authorship is implied.
Date: April 7, 1955
Creator: Wood, E.C.
Partner: UNT Libraries Government Documents Department

A critical mass study of Hood No. 5 and the filter boat cleaning and testing hood in the 234-5 Building

Description: Calculations are presented for the estimated minimum critical mass of Pu that would be allowed in Hood No. 5 and the filter boat cleaning and testing hood of the Plutonium Finishing Plant. Hood No. 5 contains the overflow tank and two vacuum traps. Limits are given for several pile exposure rates. These limits do not allow the approach of personnel into Hood No. 5 when these vessels are full.
Date: July 14, 1955
Creator: Ketzlach, N.
Partner: UNT Libraries Government Documents Department