The National Laboratories are run under contract by private bodies: Argonne National Laboratory (ANL), run by University of Chicago; Los

Alamos Scientific Laboratory (LASL), run by University of California; Oak Ridge National Laboratory (ORNL), run by Union Carbide; Brook — haven National Laboratory; and Savannah River National Laboratory.

They are almost wholly financed by the AEC and indirectly organized by the AEC. Their main function is to engage in basic research and code de­velopment, using facilities which are too large and expensive for private industry to operate.

Argonne National Laboratory and its facilities in the National Reactor Testing Station (NRTS), Idaho, forms the major force in fast reactor development. A good deal of experimental work including EBR-II and the TREAT facility comes within the ANL orbit. The ANL code center acts as a clearing house for ANL and industry codes developed under AEC contract, available for the good of fast reactor development.

Oak Ridge National Laboratory, besides having an experimental fast reactor program devoted to both sodium facilities and gas-cooled work, runs the National Safety Information Center (NSIC). The NSIC distributes abstracts on the user’s selected safety topic to the user on a regular basis. The service, operated as a totally computerized system, is free to any

Fig. 6.3. National Safety Information Center abstract card.

engineer engaged in safety work. It is very comprehensive. Figure 6.3 shows an example of an abstract card produced by the NSIC and Fig. 6.4 shows an example of a page from an NSIC-produced listing of articles which have appeared in the journal Nuclear Safety, also issued by ORNL [see general references, Chapter 1].

08-5-2-455 CABRI — A TEST REACTOR FOR SAFETY STUDIES MILLOT, J. P.

CADARACHE NUCLEAR RESEARCH CENTER, FRANCE

THE SAFETY OF SWIMMING-POOL REACTORS HAS BEEN THOROUGHLY STUD IEO IN FRANCE. THE CABRI REACTOR, DESIGNED FOR THAT PURPOSE, WAS INITIALLY USED TO INVESTIGATE REACTIVITY ACCIDENTS. LOSS-OF-COOLANT-FLOW ACCIDENTS WERE ALSO STUDIED,

AND THIS WORK IS BEING EXTENDED WITH A NEW FACILITY THAT BECAME OPERATIONAL THIS YEAR, CABRI PUISSANCE. A PROGRAM OF FUEL TESTING FOR BOTH WATER AND FAST REACTORS, INVOLVING CABRI AS A CAPSULE-DRIVER CORE, IS BEING ESTABLISHED.

08-5-2-461 RADIATION DAMAGE TO PRESSURE VESSEL STEELS WECHSLER, M. S.

OAK RIDGE NATIONAL LABORATORY, OAK RIDGEt TENNESSEE

THE EFFECTS OF NEUTRON IRRADIATION ON THE PROPERTIES OF STRUCTURAL MATERIALS ARE OF SIGNIFICANT INTEREST TO THE REACTOR INDUSTRY BECAUSE OF THEIR POSSIBLE RELATION TO THE INTEGRITY OF REACTOR PRESSURE VESSELS. IRRADIATION EFFECTS ON PRESSURE-VESSEL STEELS ARE CONSIDERED. THE VARIABLES THAT AFFECT THE NIL CONDUCTIVITY TRANSITION TEMPERATURE ARE SEPARATED INTO TWO CATEGORIES — MATERIALS VARIABLES AND RADIATION VARIABLES. AMONG THE MATERIALS VARIABLES CONSIDERED ARE CHEMICAL IMPURITIES (SUCH AS BORON,. CARBON, AND NITROGEN), GRAIN SIZEt AND METALLURGICAL STRUCTURE ASSOCIATED WITH WELDING. THE RADIATION VARIABLES INCLUDE DOSE, DOSE RATE, NEUTRON SPECTRA, AND IRRADIATION TEMPERATURE. OTHER AREAS REVIEWED INCLUDE POSSIBLE APPLICATIONS (AND LIMITATIONS) OF FRACTURE MECHANICS TO THE PROBLEM OF RAOIATION EMBRITTLEMENT IN PRESSURE-VESSEL STEELS AND THE EFFECTS OF IRRADIATION ON THE FATIGUE STRENGTH OF STEELS.

08-5-3-470 RELIABILITY ANALYSIS OF ENGINEERED SAFEGUARDS GARRICK, B. J. ♦ GEKLER, W. C.

HOLMES AND NARVER, INC., LOS ANGELES, CALIFORNIA THE REQUIREMENTS FOR ASSESSING RELIABILITY IN EN SAFEGUARDS ARE REVIEWED, AN EXAMPLE OF RELIABILI A TYPICAL EMERGENCY CORE-COOLING SYSTEM IS GIVEN PRESENTED FOR USE OF RELIABILITY TECHNIQUES IN N SAFETY ANALYSIS. RELIABILITY CAN NOW BE USEO IN NUCLEAR SAFETY, BUT, BEFORE IT CAN BE A TRULY EF CERTAIN ADDITIONAL REQUIREMENTS MUST BE MET. REL PROBABILISTIC MODELS MUST BE DEVELOPED, AND REAL STATISTICAL DATA ARE NEEDED. TECHNIQUES EXIST FO RELEVANT MODELS. STATISTICAL DATA NOW DERIVE LAR OPERATING EXPERIENCE. THESE DATA, ALTHOUGH OF SO IMPROVED TO PERMIT CONFIDENCE IN MATHMATICAL EVA RELIABILITY. THE IMPORTANCE OF RELIABILITY LIES NUMBERS GENERATED AS IT DOES IN REQUIRING SYSTEM OF SAFETY ANALYSIS AND TESTING AND ACCIDENT-PREV PROGRAMS.

08-5-3-479 RELIABILITY ESTIMATES AND REACTOR SAFETY SYSTEM OPERATION SCOTT, R. L.

OAK RIDGE NATIONAL LABORATORY, OAK RIDGEt TENNESSEE

TWO PAPERS PRESENTED AT THE WINTER 1966 MEETING OF THE AMERICAN NUCLEAR SOCIETY ARE SUMMARIZED. ONE PAPER DISCUSSES DATA OBTAINED FROM FIVE POWER-REACTOR OPERATING AGENCIES IN ORDER TO MAKE RELIABILITY ESTIMATES. SINCE THE DATA WERE OBTAINEO FOR PURPOSES OTHER THAN RELIABILITY PREDICTIONt IT HAS LIMITED APPLICABILITY. RELIABILITY ESTIMATES ARE PRESENTED FOR EMERGENCY CORE-COOLING SYSTEMSt EMERGENCY POWER SYSTEMSt AND SECONDARY SHUTDOWN SYSTEMS. THE OTHER PAPER DISCUSSES REACTOR-SAFETY SYSTEM OPERATION DATA OBTAINED FROM THE SAME FIVE POWER-GENERATING PLANTS. THE COMPOSITE DATA INDICATE THAT REAL AND SPURIOUS SCRAM RATES ARE EACH BETWEEN FIVE AND SIX PER YEAR. MOST OF THESE SCRAMS OCCURRED AT LOW REACTOR POWER AND PROOUCED ONLY ONE OR TWO POWER-GENE RAT I ON OUTAGES PER YEAR.

Fig. 6.4. Page of National Safety Information Center listing of articles which have appeared in Nuclear Safety.

References and abstracts are located by certain key words. Thus, by specifying a number of key words such as reactor, fast, melt-down, accident — analysis, the user is able to specify a very well defined topic of his interest on which he would like information.

In addition, ORNL produces survey documents on selected technical topics such as missile protection, nuclear reactors and earthquakes, etc. This service is also an AEC-financed information service.