None

None

The combination of a large reactor core with a sufficiently high flux level results in complications of the dynamic behavior of the core due to asymmetric xenon and temperature instabilities. In both phenomena, a local increase in neutron flux causes an increase in neutron multiplication in a surrounding zone. We consider here the instability associated with the delayed production of Xe{sup l35}, an isotope with a very high thermal neutron cross section. In order for this instability to result in flux oscillations., the neutron production in one part of the core must be independent of conditions in another part. Said in another way this means that the neutron migration area must be considerably less than the square of any core dimensions. Thus, even though the total reactor power is held constant, it may be possible that neutron leakage from one part of the reactor to another is insufficient to counteract the tendency of the local flux to continue increasing in the part of the reactor where it is already high. An asymmetry in the distribution of reactor power may thus tend to increase in After some initial confusion., it has been found that close monitoring of the local power trends …

The costs of pot calcination of Purex and Thorex wastes were calculated. The wastes were assumed produced by a plant processing 1500 ton/year of U converter fuel at a burnup of 10,000 Mwd/ton and 270 ton/year of Th converter fuel at 20,000 Mwd/ton. Costs were calculated for processing Purex waste in acidic and reacidified forms and for processing Thorex wastes in acidic and reacidified forms and with constituents added for producing an acidic Thorex glass. Calcination vessel designs were right circular cylinders similar to those used in engineering development studies. Costs were calculated for processing in 6-, 12-, and 24-in.-dia vessels with a fixed length of 10 ft. Vessel costs used, based on estimates from private industry, were calculated for wastes decayed 120 days and 1, 3, 10, and 30 years after reactor discharge prior to calcination. Aging had negligible effect on costs, except as it permitted larger diameter vessels to be used, because vessel and operating costs were much larger than capital costs in all cases. The lowest cost was 0.87 x 10/sup -2/ mill/kwh/sub e/ for processing acidic Purex and Thorex wastes in 24-in.-dia vessels, and the highest was 5.0 x 10/sup -2/ mill/kwh/sub e/ for processing reacidified …

A 4-ft-long boiler was successfully fabricated for a boiling-liquid- metal loop. This involved brazing massive copper blocks betwveen a thin-walled stainless steel tube and a thin-gage stainless steel jacket. Essentially 100% bonding to the tube was required while a few areas of nonbond were permitted in the copper-to-jacket braze. The brazing was accomplished in vacuum, using an electrolessdeposited nickel-phosphorus alloy and slurry-deposited nickel-chromium --phosphorus alloy. Bonding was better than 99% complete in the copper-to-tube braze and was approx 92% complete in the copper-to-jacket braze. (auth)

Electric Co. Hanford Atomic Products Operation, RichIrradiation experiments are being performed in the GETR to determine the long-term stability of EGCR graphite. The results of two experiments with a peak total exposure equivalent to 38,000 MWD/AT in the EGCR are available. This is approximately one half the expected lifetime exposure of the EGCR. The H-3-l and H-3-2 capsules contained samples of EGCR graphite and reference CSF graphite. Both capsules were identical and were irradiated sequentially to increasc the dose on the samples. Sixty per cent of the samples irradiated in the H-3-l capsulc were reirradiated in the H-3-2 capsule. The capsule design utilized the gamma heating of the GETR E-7 facility to achieve desired operating temperatures. The temperatures on inter- and intra-cycle irradiation were constant within 25 f C. Samples were irradiated in a temperature range of 475 to 850 f C. The graphite tested was obtained from a fullsize cross section of an EGCR block. Irradiation results did not indicate differences in contraction of parallel samples chosen at several positions within the block. Contraction rates of the graphites tested were found to depend on the temperature of irradiation. Estimates of the contraction rates of parallel samples of EGCR graphite …

This is the monthly report for the Hanford Laboratories Operation September 1961. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, operations research and synthesis, programming, and radiation protection operation are discussed.

This document details activities of the irradiation processing department during the month of September, 1961. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; and NPR Project.

The purpose of the test described in this report is to set an upper limit on shadowing of the DR Reactor vertical safety system control strength by residual boron-steel balls in the graphite stack. Such an experimental determination should permit less stringent total control requirements than those based presently on necessarily conservative assumptions.

Laboratory heat transfer experiments were conduced to investigate fuel element temperatures which could result from coolant flow loss following a failure of the inlet piping to a process tube at a B, D, F, DR, or H reactor. The results are reported herein. Failure of the inlet coolant piping between the front header and the process tube on a reactor would stop the normal flow of cooling water to the fuel elements. Such a failure should immediately initiate a reactor shutdown, but the only means of removing the heat released during the post-shutdown period would be by reverse flow of hot water from the rear cross header. The subject experiments were conducted to determine what rear header pressure would be required to achieve adequate cooling of a BDF type reactor fuel assembly following such a piping rupture. Experimental studies were previously reported concerning failure of inlet piping to a K reactor geometry. The analytical techniques and experimental procedures used previously were also used in the present experiments.

None

None

ABS>The results are presented of critical water height measurements made on close-packed lattices of Spert III, highly enriched, plate-type, stainless- steel-clad fuel elements. Experiments were conducted with cores containing no control rods and with cores containing a single, fully-inserted control rod. The "clean critical" data obtained in these experiments were used to test the validity of various aspects of a four-group, diffusion theory analysis of the full scale Spert III reactor. The results of the analyses of the rod-free and single-rodded critical lattices show that for such stainless steel cores k/sub eff/ can be calculated to within 1% DELTA k and that the Spert III control rod worth is calculable to a few tenths % DELTA k. (auth)

It is now generally known by those working on Tory II-A that the average maximum fuel element wall temperature obtained during the Intermediate Power Test was somewhat higher than the 2250 degrees F design value. The purpose of this report is to explain how this occurred.

The objective of this supplement is to review the basis, justification and procedure so that this production test may be used permitting a determination of minimum downtime, when advisable for improving moderator coefficient data as function of exposure.

A flowsheet, based on laboratory-scale data, is presented for oxyhydrochlorination of 90% uranium --10% molyhdenum alloy with 15% HCl in air at 4OO deg C in 18 hr. Up to 90% of the molybdenum is volatilized during oxyhydrochlorination and another 3 to 6% is removed by a 2-hr treatment with pure hydrogen chloride at 400 deg C. Residual chloride is removed by a 4-hr treatment with moist air at 400 deg C, and the product uranium oxide is dissolved in 4M nitric acid to yield a stable solvent extraction feed solution of 1M uranium, 0.017M molybdenum, 175 ppm chloride, and 1.7M nitric acid. The stainless steel cladding of the original fuel would be removed mechanically and the core recanned in aluminum prior to transfer to the core processing facility. The aluminum can would be removed by hydrochlorination prior to core treatment. (auth)

The objective of this production test is to evaluate the behavior of large diameter tubular UO{sub 2} fuel elements during high temperature irradiation. Eighteen inch long tubular UO{sub 2} fuel elements 1.804 inch OD, 0.544 inch ID, in 0.060 inch Zircaloy-2 jackets will be irradiated in the KER loops either alone or in conjunction with other tests to exposures up to 3500 MWD/T of contained uranium.

K area emergency water backup studies have been hampered by poor data on flow through the reactor under various emergency conditions. Various tests have been run where emergency conditions have been simulated and flow measurements attempted. In all previous tests, the accuracy of the flow measurements have been questionable. Flow from the high-pressure crosstie can be measured by an orifice in the crosstie, but there has not been any method of measuring the service water contribution to total reactor flow under simulated emergency conditions. One method of measuring the total reactor flow regardless of its source is to determine the relationship between total flow through the reactor and the bottom of riser pressure. After this relationship has been determined for the flow range of interest, then flow to the reactor can be determined by reading bottom of riser pressure (BORP) and converting that to flow. The objective of this production test was to obtain the relationship between BORP and total reactor flow in the range of 10,000 gpm to 25,000 gpm. An additional objective of this test was to check the accuracy of the No. 2 pump discharge venturi.

None

Earlier this year we briefly discussed the potential hazard of incurring an inadvertent uranium fuel element fire during discharge. This letter will provide data which will be of assistance to you in assessing the potential hazard, and in establishing charge-discharge procedures to minimize the probability of an irradiated fuel element lodged in the discharge area reaching aluminum jacket melting temperature without detection.

A nonnuclear explosion involving an evaporator occurred in a shielded cell in the Radiochemical Processing Pilot Plant at Oak Ridge National Laboratory on Nov. 20, 1959. Plutonium was released from the processing cell, probably as an aerosol of fine particles of plutonium oxide. It is probable that this evaporator system had accumulated -1100 g of nitric acid-insoluble plutonium in the steam stripper packing; the explosion released an estimated 150 g inside Cell 6, with about 135 g in the evaporator subcell, and about 15 g in the larger main cell. No radioactive material was released from the ventilation stacks; no contamination of grounds and facilities occurred outside of a relatively small area of OaK Ridge National Laboratory immediately adjacent to the explosion. No one was injured by the explosion, and no one received more than 2% of a lifetime body burden of plutonium or an overexposure to sources of ionizing radiation either at the time of the incident or daring subsequent cleanup operations. The explosion is considerdd to be the result of rapid reaction of nitrated organic compounds formed by the inadvertent nitration of about 14 liters of a proprietary decontaminating reagent. In cleanup the contamination was bonded to the …

In coolant system such as NPR's, the coolant activity level increase with each succeeding pass through the reactor flux until a saturation limit is reached. Therefore, the activity level of the NPR coolant system will be much higher than that of the old reactor once-through systems. This report is the determination of the specific activities (disintegrations/cc{center dot}sec) of the various coolant impurities which determine the total activity of the coolant system. 10 refs., 13 figs., 2 tabs.

In relation to evaluation of the block discharge mode of operation, data for comparison of fuel rupture performance of the old reactors have been assembled, and are transmitted herein. We have analyzed these data and the results are also transmitted in this report. Based on analysis of rupture experience from March through December, 1960, adjusted for differences in power, temperature, exposure, and throughput, there are no significant differences in fuel rupture performance between the old reactors. This statement does not imply that there are no differences in economic performance between reactors. Differences of this nature cannot be judged solely from the data of this report.

Four methods of rebunching protons within Brooknaven's AGS are discussed. The first method involving switching off the old r-f and switching on the new r-f simultaneously with the new r-f increasing adiabatically seems impractical. Two other methods utilize gradual removal of the old r-f voltage and introduction of the new r-f voltage, or the reverse. Removal of the old r-f voltage followed by introduction of the new seems to give the best results. Several phase diagrams are included. (D.C.W.)