605 Matching Results

Search Results

Advanced search parameters have been applied.

REACTOR PLANT SELF-ACTUATED RELIEF VALVE OPERATION. CORE 1, SEED 1. Test Results T-643700

Description: The relief valves that were tested included four each in the reactor coolant loop and in the purification system. Test results are tabulated which indicated that all valves operated at or witai the prescribed limits of 2530 plus or minus 50 psig in the "as left" condition. Most of the valves had to be popped and adjusted several times befors the valves would pop at a pressure within the prescribed limitations. Some of the difficulties encountered during the testing are discussed. (B.O.G.)
Date: October 31, 1961
Partner: UNT Libraries Government Documents Department

MONTHLY PROGRESS REPORT FOR THE PERIOD AUGUST 1 TO 31, 1958

Description: plants ln England and France. With the increasing de output of given designs and probably allow operation at higher polymer contents than orignally foreseen, thereby reducing the make-up requirements. The physical characteristics of the OMRE such as critical loading, temperature coefficient, and general stability appeared to be close to the predicted values. Radiation levels in the primary circuit area during full power operation appear to be so low that maintenance is possible during operation. The reactor has been run for a full month at 30% polymer concentration and is, at the time of this writing, brought to a still higher steady state percentage of breakdown products ln the coolant stream. No evidence whatsoever of fouling or precipitation has been observed. The reactor behaves in a routine manner in all respects and invites immediate application of the OMR principle to reactors for large scale ceniral stations. Final design on one 11.4 Mwe unit for the city of Piqua, Ohio, has now stanted. A short description is given of OMR power reactors. The use of magnetic jack mechanisms for control and safety rods provides a reactors top shield without penetrations, as well as an unpenetrated lower core vessel, still avoiding any interference from the control rods during fuel changing. The new finned-plate fuel element is mentioned as well as the use of a liquid pressurizing pump instead of nitrogen gas pressurization. It is conservatively predicted that the cost of organic liquid make-up for these designs will not contribute more than one half to one mill per kwh to the total power cost. In case operation at higher polymer concentrations appears practicable, this figure may even be lower. More detailed pricing informntion available now, has shown that the original cost estimate of around 0 per kw installed for a 150 Mwe plant can ...
Date: September 20, 1958
Creator: Garbe, R.W. & Walchli, H.E.
Partner: UNT Libraries Government Documents Department

VALVE OPERATING SYSTEM SYSTEM DESCRIPTION NO. 10

Description: A description is given of the valve operating system for the PWR. The valves served by this system are a component pant of the following systems: reactor coolant system; pressurizer and pressure relief system; coolant discharge and vent system; and failed element detection and location system. (W.L.H.)
Date: May 1, 1957
Creator: None
Partner: UNT Libraries Government Documents Department

TRIP TO ALCO PRODUCTS, KAPL, COMBUSTION ENGINEERING CO. AND FT. BELVOIR (APPR). RE: COOLING WATER ACTIVITY BUILDUP IN PRESSURIZED WATER REACTORS

Description: Several uncertainties associated with predicting activity levels in water-cooled reactors are discussed in connection with development of the Nuclear Merchant Ship Reactor (Maritime Pressurized Water Reactor). (T.R.H.)
Date: August 1, 1958
Creator: Culver, H.N.
Partner: UNT Libraries Government Documents Department

DESIGN CRITERIA FOR A FAST FLUX LIQUID METAL LOOP IN THE ADVANCED TEST REACTOR

Description: Design criteria are presented for liquid-metal loop facilities for the Advanced Test Reactor. It was concluded that achievement of the design objectives could not be predicted with any high degree of confidence when utilizing the package loop concept based on the design philosophy of the PW-19 loop. (M.C.G.)
Date: July 1, 1962
Creator: None
Partner: UNT Libraries Government Documents Department

Mathematical modeling of a Fermilab helium liquefier coldbox

Description: Fermilab Central Helium Liquefier (CHL) facility is operated 24 hours-a-day to supply 4.6{degrees}K for the Fermilab Tevatron superconducting proton-antiproton collider Ring and to recover warm return gases. The centerpieces of the CHL are two independent cold boxes rated at 4000 and 5400 liters/hour with LN{sub 2} precool. These coldboxes are Claude cycle and have identical heat exchangers trains, but different turbo-expanders. The Tevatron cryogenics demand for higher helium supply from CHL was the driving force to investigate an installation of an expansion engine in place of the Joule-Thompson valve. A mathematical model was developed to describe the thermo- and gas-dynamic processes for the equipment included in the helium coldbox. The model is based on a finite element approach, opposite to a global variables approach, thus providing for higher accuracy and conversion stability. Though the coefficients used in thermo- and gas-dynamic equations are unique for a given coldbox, the general approach, the equations, the methods of computations, and most of the subroutines written in FORTRAN can be readily applied to different coldboxes. The simulation results are compared against actual operating data to demonstrate applicability of the model.
Date: December 1, 1995
Creator: Geynisman, M.G. & Walker, R.J.
Partner: UNT Libraries Government Documents Department

FLOW DISTRIBUTION ACROSS THE CORE. CORE I, SEED 2, Section 2. Test Results T-550097

Description: An investigntion was conducted to determine the reactor coolant mixing charactsristics of the inlet plenum chamber. It was found that mixing in the inlet plenum was slight, and flow from a given loop is corfined mainly to its quadrant of the core. There appears to be no significant change in the flow rate through the fuel assemblies monitored with the reactor coolant introduced at various temperatures and the reactor coolant pump on either fast or slow speed. (J.R.D.)
Date: January 31, 1961
Partner: UNT Libraries Government Documents Department

PRELIMINARY HAZARDS EVALUATION OF THE ISOLATED COOLANT LOOPS IN THE HWCTR

Description: The design features of two isolated coolant loops in the Heavy Water Components Test Reactor (HWCTR) and the effect of these features on reactor safety are described. It is shown that limitations on the operating conditions, backed up by properly chosen safety systems, will permit operation of the loops without adding appreciably to the reactor hazards. (auth)
Date: July 1, 1960
Creator: Arnett, L.M.; Parkinson, T.F.; Randall, D. & Ross, C.P.
Partner: UNT Libraries Government Documents Department

THE CORROSION OF ALUMINUM ALLOYS IN THE OAK RIDGE RESEARCH REACTOR

Description: A corrosion testing program designed to estimate the potential service life of aluminum alloys used in the construction of the Oak Ridge Research Reactor (ORR) cooling systems has been in progress for over two years. The five alloys (1100, 3003, 5052, 5154, and 6061) used to the greatest extent in the reactor exhibited continuously decreasing corrosion rates since the first 500-hr inspection. Samples exposed in the core-cooling loop have shown a decrease in corrosion rate from a 2.6 mpy maximum for one group during the first 500 hr to an over-all average of less than 0.1 mpy for another group after a full year in test, with the maximum metal loss less than 0.1 mils. Results indicate that with suitable water treatment the aluminum alloys used in the ORR may be expected to give satisfactory performance for many years. Based on the generalized corrosion rates alone, 40 to 50 years of service life may be expected. However, since occasional localized corrosion has been observed (rarely), minor repairs will almost certainly be required before that time. (auth)
Date: July 1, 1961
Creator: Neumann, P.D.
Partner: UNT Libraries Government Documents Department

NUMERICAL SOLUTION OF FUEL-ELEMENT THERMAL-STRESS PROBLEMS

Description: In developing a method of numerical analysis for the solution of thermal- stress problems special emphasis was given to fuel elements with internal coolant channels. Numerical techniques ior reducing the partial differential equation system te a form suitable for numerical solution and a new iteratlve method of solving large systems of linear algebraic equations were employed. Computer codes were devised to obtain the numerical solution of the thermal-stress problems and were used to obtain numerical results for single-hole and seven-hole hexagonal elements and plate-type elements. Comparisons were made between analytical results and numerical results for the case of t:.ie simple annulus shape. (auth)
Date: February 26, 1960
Creator: Redmond, R.F.; Pollack, H.; Klickman, A.E.; Hogan, W.S.; Epstein, H.M. & Chastain, J.W.
Partner: UNT Libraries Government Documents Department

PERIODIC CALIBRATION OF TEMPERATURE SENSING ELEMENTS. CORE I, SEED 2. Test Results (T-641303-B)

Description: Tests were conducted to determine the direction and magnitude of any short- or long-term drift in core thermocouples, primary loop and pressurizer resistance thermometers, and resistance thermometers located in the secondary side of the boilers for Seed 2 at 2303 EFPH. It was found that drift in the core thermocouples is definite but random. Other findings are discussed and calibration data are presented graphically. (J.R.D.)
Date: April 14, 1961
Partner: UNT Libraries Government Documents Department

PERIODIC INTERCALIBRATION OF TEMPERATURE SENSING ELEMENTS. CORE I, SEED 2. Test Results T-641303B

Description: Calibration testing was conducted to determine the direction and magnitude of drift in core thermocouples, primary loop and pressunizer resistance thermometers. and the resistance thermometers in the secondary side of the boilers. Because of operational conditions, only those sensing elements in coolant loops A and C could be evaluated. An analysis of data from A and C loops is included. (J.R.D.)
Date: January 31, 1961
Partner: UNT Libraries Government Documents Department

EXAMINATION OF COMPONENTS FOR CRUD AND CORROSION. CORE I, SEED 2. Test Results T-612080

Description: >An examination was made to observe the extent and location of corrosion, crud deposits, and defects in components of the PWR pnimary fluid system and its auxiliaries. A gamma spectrum of the four-inch line upstream from the two pressurizer self-actuated relief valves showed the presence of Mn/sup 54/ and Co/sup 60/. From the gamma spectrum and the gross gamma activity, the Co/sup 60/ was found to be 5.03 x 10/sup 3/ dpm/mg or about 70 per cert of the gross gnmma activity. (J.R.D.)
Date: January 20, 1961
Partner: UNT Libraries Government Documents Department

CORE LEVITATION IN THE EGCR IN CASE OF MAIN COOLANT PIPE FAILURE

Description: Results of an analysis to determine the extent of displacement of the EGCR core due to blowdown in case of several postulated hot main gas coolant pipe failures are summarized. Results show that the core will be damaged for ary hot pipe double-ended failure. Excepting the improbable case of no coolant flow existing prior to the break, the core will be damaged for any hot pipe fracture exposing a total flow area to the atmosphere equal to that of one pipe. Smaller breaks will probably be safe in this respect. (auth)
Date: August 4, 1959
Creator: Fontana, M.H.
Partner: UNT Libraries Government Documents Department

Hydrogen Distribution and Leak Rate From the Reactor Coolant System. Test Results, RNI-31 A. Core I, Seed 1

Description: No conclusions could be made concerning tbe distribution of hydrogen injected into the reactor coolant because the samples were taken only from the purification loop through which tae hydrogen was added. No comparison of the uniformity of distribution obtained by injecting hydrogen through either the AC or the BD purification loop could be made. Comparative data were not taken until 9 days had elapsed since injecting hydrogen and the concentration was near tbe lower limit (15 cc/kg). The rates of hydrogen loss durirg the test were 5.9 cc/ kg/day for hydrogen injected through the AC purification loop, and 4.1 cc/kg/day for hydrogen injected through the BD purification loop. (auth)
Date: June 1, 1960
Partner: UNT Libraries Government Documents Department

PERIODIC REACTOR PLANT LEAK RATE TEST. CORE 1, SEED 2. Test Results

Description: An investigation was made to determine the magnitude and location of the reactor coolant system leakage of the Shippingport Power Beactor. The "total" plant leakage was indicated by the pressurizer level change since no make-up was added during the test. The "accounted for" leakage was indicated by the flash tank level change and the blowoff tank level change. The relief valves from the pressurizer and reactor were checked for leakage individrually and a combined leakage of 2.84 gal/hr was obtainsd for the 6 valves tested. The system was found to have a total leak rate of 5.65 gal/hr of which 1.17 gal/hr was "accounted for." (M.C.G.)
Date: January 20, 1961
Partner: UNT Libraries Government Documents Department

FLOW DISTRIBUTION ACROSS THE CORE. CORE I, SEED 2. Section 1. Test Results T-550097

Description: An investigntion was conducted to determine flow distribution characteristics within the core and determine any relative shift in flow between the seed and the four blanket regions. The highest average flow measured throngh the blanket regions, for each loop arrangement, was established, however the flow in other regions was not consistent enough to establish a pattern. (J.R.D.)
Date: January 31, 1961
Partner: UNT Libraries Government Documents Department