Eight Zircaloy-2 jacketed, natural uranium seven-rod cluster elements were irradiated in a KER loop to determine flow channel temperature characteristics. One of the elements, which had 200 MWD/T exposure, was sent to the Radiometallurgy Laboratory for examination in April 1959. An outside rod of the cluster was sectioned and examined metallographically. No cracks or flaws were observed in the uranium cladding or bonding.

Defect testing of the irradiated specimens will provide kinetic autoclave defect test data on the behavior and performance of coextruded Zr-2 clad fuel failures; this should reveal the effects of exposure on the coextruded U-Zr-2 bond.

The objectives of this supplement are to test an improved prototypic slug column displacement winch and associated remote control equipment and to demonstrate the feasibility of exerting control over the front-to-rear neutron flux changes as an aid in controlling reactivity cycling. An array of poison displacement columns up to twelve in number is intended, each charge consisting of a center section of I and E natural uranium slugs with mint pieces on each end. Initially six such columns will be charged with the remaining six being added later if required for proper control. The initial testing of the slug column displacement winch has established that the mechanics of the system are practicable and has indicated that the system would be feasible. The next phase of this test is to demonstrate the feasibility of controlling the front-to-rear flux changes over the entire reactor and to verify the improvement in equipment design. To be effective in controlling front-to-rear flux changes, there should be poison displacement tubes available in each quadrant of the reactor. It is believed that a maximum of twelve such tubes will be sufficient to show effective control with six of these being charged initially and the others added if …

The purpose of this test was (1) to determine the conversion ratio of alternated enriched uranium fuel segments and lithium aluminum target slugs, and (2) to determine the stability of solid and cored enriched uranium with this type of load. The test consisted of an irradiation of two twelve-tube pile charges, each of them forming a sixteen tube square without corners. The first twelve tubes were a ``striped`` load containing solid fuel elements enriched to .95% U-235 alternated with lithium-aluminum target slugs. The fuel elements of the second array were 1/2 inch cored, but canned with the core plugged to give the same O.D. and external appearance as a solid element. To determine fuel element stability, the irradiation was essentially a run-to-rupture test. The central four-tubes of each square were designated for special extraction after the irradiation to determine conversion ratio. Neither solid nor cored enriched uranium of metal quality comparable to that used in this test has adequate stability to reach a routine goal exposure of more than 500 MWD/T at the specific powers of the test, 70--80 KW/foot. From post-irradiation measurements of diameter growth of the cored slugs, it might be concluded that many of the slugs were …

The possibility of overboring in existing Hanford reactors to increase slug size has been suggested as a means of increasing production. The effect on control system strength of such a modification is, of course, one of the factors which must be considered. This report presents the results of a study requested by the Reactor Physics Unit of the local buckling of control rods as a function of the migration area. The migration area varies for different lattice make-ups, but due to the complexity, calculation of T and L{sup 2} for changes in slug and hole size was not done here; instead, the range of values given by the ``Giant`` Program was assumed.

Objective of this production test is to obtain irradiation experience wit the hot-headed closure for co-extruded Zircaloy-2 jacketed rod. Seven Zircaloy-2 jacketed natural uranium seven-rod cluster elements with hot-headed end closures and spark machined end supports will be irradiated in the KER Loops to an exposure of 2000 MWD/T in pH 8--11 coolant.

The purpose of this report is to outline the preliminary design and provide justification for the installation of crossunder lines at B, D, DR, F, and H Reactors.

A metallurgical bond can be effected between a Zircaloy-2 jacket and a thin walled uranium tube by treatment in a high temperature, high pressure gas autoclave. Bonding is independent of core history, provided that all surfaces are free from contamination. Gas pressure bonding causes the fuel element jacket to conform to the core, which promotes bonding on all surfaces including the ends. Grain growth occurred in some instances between the jacket and end cap interface. A three phase system (alpha, delta and epsilon) was identified across the interface at the bonding conditions of 845 C (1500 F), 10,000 psi and a time of four hours. Four structures were observed; these were epsilon, epsilon plus delta, delta, and alpha plus delta. Microhardness tests were used to identify the various phases. Also, thickness measurements were made of each component in the interface. The normal pressure bonded interface consists mainly of epsilon and delta phases. A cast bonded sample, subjected to the pressure bonding treatment, consists predominantly of alpha plus delta material while an interface bonded during casting is primarily alpha plus delta phases and epsilon phase.

The objective of this supplement is to authorize a change in the panellit trip range from 25--75 psi to 5--95 psi. The test hole facility consists of two concentric aluminum tubes which extend from the front face to the rear face of the reactor. The ID of the inner tube is 2--7/8 inch. Water from one crossheader supplies the annulus, water from another crossheader supplies the inner tube. The three-foot-long, .570 inch OD fuel element is centered in a 40-inch long aluminum holder which has an ID of 1.380 inch and an OD of 2.800 inch. The panellit gage which monitors the flow to the inner tube fluctuates to such an extent during start-up that on two occasions the reactor was scrammed. During equilibrium operation the panellit gage reading remains stable. A possible explanation of this behavior is that during start-up aluminum spacers which are in the inner tube as part of the test charge chatter and cause variations in the water path through the tube. It is further surmised that at equilibrium operation the pressure drop across the column in the tube is sufficient to suppress the chattering. It is concluded that extending the trip range to 5--95 psi …

This report discusses the service life expectancy of the 105-N Reactor, Steam generation transients following a scram, and estimated number of outages.

Inconel can be used for thermal neutron flux measurements by means of its cobalt impurity or its chromium constituent where conventional monitors are unsuitable. The use of cobalt should also be applicable to other nickel alloys. Discriminatory counting is required. (auth)

Liquid and dry processes suituble for the purification of gases by the removal of CO/sub 2/, H/sub 2/O, CO, H/sub 2/, and hydrocarbons are discussed. Recommendations are given for specific processes io be included in a "dry" (no liquid absorbents or chemicals used) purification system for the hellum coolant of the EGCR The recommended processes include (1) a catalytic converter for the oxidation of CO, H/sub 2/, and hydrocarbons to CO/sub 2/ and H/sub 2/O, (2) cooler-condensors for the removal of the bulk of the R/sub 2/O, (3) silica gel adsorbers to complete the removal of H/sub 2/O, and (4) Linde Molecular Sieve adsorbers for the removal of CO/sub 2/. No provisions are included for the planned removal of radioactive gases or particulates. (auth)

Twelve irradiated specimens of thorium -11 wt.% U were examined. The specimens were fabricated by induction melting and casting in graphite and cold swaging to about 42% reduction in area. The irradiations were conducted in the MTR in capsules equipped with thermocouples. Six specimcns were irradiated to burnups ranging from 0.5 to 0.6 total at.% at average central corc temperatures ranging from 1070 to 1250 F. Three spccimens exbibited sevcre swelling or decrepitation and three appeared to be in relatively good condition. The density of these specimens decreased from 4.9 to 9.9%. The remaining six specimens were irradiated to burnups ranging from 0.9 to 1.5 total at.% at average central core temperatures ranging from 970 to 1100 F. These specimens were in relatively good condition, except for three that had swollen sevcrely at one end. Density decreases ranging from 2.4 to 3.8% were determined for these specimens. Swelling of all specimens appeared to be a linear function of burnup to the highest level studied (l.5 total at.%) and at temperatures below about 1100 F. Swelling increased significantly above 1100 F, even at burnups as low as 0.2 at.%. Fission-gas losses averaged about 0.5% for sound specimens after burnups of 1.2 …

In the fabrication of the steam generator on APPR-1A it was considered necessary to roll the Type 430M tubes into carbon steel tubesheets to ASTM Specification A350-Grade LF-1, modifled with 1.66% nickel; and weld the tube ends to the stainless steel overlay previously applied to the tubesheet. The rolled joint was a necessary precaution to prevent secondary water, that might contain chlorides, from contacting the stainless steel weld joining the tubes to the tubesheets. The welded joint provided the mechanical strength for attaching the tubes to the tubesheets. A laboratory program was conducted, therefore, to develop practicable procedures for welding the Type 430M tubes to the stainless steel overlay; as well as to assure that the tubes could satisfactorily be rolled to the tubesheets. Automatic and manual tungstenare welding procedures were developed that were capable of consistently providing an austenitic weld having a strength exceeding that of the heat affected zone or the unaffected tube itself. Type 430M tubes in the asreceived, and softened conditions were rolled into prototype test units under various conditions of rolling. It was concluded that the Type 450M tubes in the as-received condition could be satisfactorily rolled into the A360Grade LF-1 tubesheet and be tlght …

Radioactive strontium can serve as a tracer to gain information concerning calcium metabolism in human subjects. Gamma-emitting Sr{sup 85} is used rather than the much more hazardous, beta-emitting Sr{sup 89} and Sr{sup 90}. (ca{sup 47} -- the ideal tracer for normal calcium -- is quite expensive and difficult to procure.) Very significant information may be obtained merely by measuring and recording the changes in radioactivity in various body areas during the first hour after intravenous injection of the bone-seeking radioisotope. This is accomplished by placing a lead-shielded gamma-scintillation detector in contact with the skin over the sites of interest and recording the activities on a scaler or ratemeter. The activity versus time curves so obtained are called radioisotope osteograms. Data were presented which indicated that Sr{sup 85} osteograms for patients afflicted with osteoporosis, Paget`s disease, tumor metastases to bone, and possibly multiple myeloma, differ significantly from those obtained from subjects with no skeletal abnormalities. Some interpretations of these deviations were discussed. The value of conducting double-tracer tests (e.g. -- Sr{sup 85} plus radio-iodinated serum albumin) was demonstrated, and correlations with excretion data were made. With further refinements the technique may ultimately become useful for certain diagnostic problems in the clinic …

An analysis of the temperature structure in the GCR2 fuel elements and coolant stream at the position where the maximum fuel element surface temperature exists was presented in a previous paper (CF-58-5-97). It was felt that the peripheral temperature variation existing on the fuel rods as brought out in that analysis was due mainly to the poor flow distribution of the coolant with respect to the heat flux. The most expedient way to alleviate this situation without changing the fuel element assembly itself is to scallop the channel so that it becomes, in effect, seven channels merged together to form a single passage. This geometry was studied by use of the IBM 704 digital computer in much the same way the earlier problems were investigated. Since the heat transfer coefficient depends much more strongly on free flow area than on equivalent diameter, the free flow area of the model reported on was made equal to that of the 3.25 in. circular channel so that a comparison of the results of the two configurations would be meaningful.(auth)

Plutonium, thermally cycled through the low-temperature allotropic transformations, exhibited extensive physical damage. The physical damage was greater than that reported for any other metal or alloy. The extent of physical damage was determined by measurements of fluid displacement and the dimensional changes. The internal porosity was examined metallographically. Physical damage varied considerably, depending on the cycling conditions and the characteristics of the metal (particularly inclusions and casting imperfections). Observations showed that increasing the specimen size produced a greater decrease in density as a function of the number of cycles. The degree of void formation was directly proportional to volume change associated with the phase transformation. Other variables such as the length-diameter ratio (constant diameter) had little or no influence on the amount of physical damage. The dimensional growth was both radial and longitudinal. The tensile strength and the yield strength of tensile specimens cycled ten times between the gamnna and alpha phases were decreased to an average of 30,000 and 26,000 psi, respectively. (auth)

7 = 9 9 9 9 7 7 7 = 9 9 9 95 : > @ 9 ; 5 8 @ = K : . ighpurity Nb deformed by impact or slow compression at - 196 deg C. An apparent phase transformation was detected in high- purity Ga deformed at 4.2 deg K. The specific heat of the group IV-A metals and alloys of Zr-In and Zr-Sn were measured from 1.2 to 4.5 deg K. In the Zr-rich portion of the Zr-Ga phase diagram, the alpha / beta phase boundaries of Zr are depressed by additions of Ga and the beta phase decomposes by a eutectoid reaction. The Cd pressures of alpha - and beta - Zr alloys containing 1 to 11% Cd were measured between 1090 and 1325 deg K. Crystal structures of several unreported transition-metal fluorides, rare-earth hydrides and nitrides were determined. Progress in the study of phase transitions in beta -quenched Zr-Nb alloys aged below the eutectoid temperature is reported. A high-temperature investigation of the order-disorder phase transition of a Cu31 at.% Au alloy has revealed an intermediate periodic antiphase condition. A previously described x- raydiffraction technique for the measurement of the thickness and strain of …

A general numerical procedure, called the discrete S/sub n/ method, for solving the neutron transport equation is described. The main topics relate to the derivation of suitable difference equations, and to the problem of solving these, while maintaining generality, accuracy, and reasonable computing speed. A few comparisons with other methods are made. (auth)

None