ML19259D072

Summary of 790628 Meeting W/Ga Co in Bethesda,Md Re Revised Design of Gas Cooled Fast Breeder

ML19259D072
Person / Time
Site:	Peach Bottom, General Atomics, 05000535
Issue date:	08/02/1979
From:	Williams P Office of Nuclear Reactor Regulation
To:	Varga S Office of Nuclear Reactor Regulation
References
NUDOCS 7908280801
Download: ML19259D072 (25)

Text

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1. o UNITED STATES S.Y 3.. r l',i NUCLEAR REGULATORY COMMISSION

j. E WASHINGTON, D. C. 20555

';&;.,.sJ,j y MEMORANDUM FOR: Steven A. Varga, Acting Assistant Director for Light Water Reactors Division of Project Management FROM: Peter M. Williams, Project Manager Division of Project Management

SUBJECT:

REPORT OF MEETING ON UPFLOW DESIGN FOR THE NASAP-GCFR I Summary A meeting was held on June 28, 1979 in Bethesda under the auspicies of the NASAP* review in which General Atomic made its initial presen-tation of its revised design of the Gas Cooled Fast Breeder Reactor (GCFR). In the revised design the normal flow direction is upward and einergency core cooling under pressurized conditions could be achieved by natural convection. While this design results in changes in the core support design, many features of the earlier GCFR design are retained or features of current HTGR designs adopted. Analyses usingtheRATSAMcodeshowedpeakcladtemperaturestobelessthan 1500 F using natural convection for emergency core cooling. The desigr, changes and the natural convective cooling analysis were presented in principal with many details remaining to be discussed. An amendment to Volume V of the NASAP Preliminary Safety and Environ-mental Information Document describing this material will be submitted shortly. A list of attendees is attached. A second enclosure provides selected copies of the Vu-graphs presented. NASAP Nonproliferation Alternative Systems Assessment Program n'9 029 200eajloSol.

Steven A. Varga II Design Changes Design Studies for an upflow core cooling arrangement for the GCFR have been underway at General Atomic for over a year, as a portion of a co-operative arrangement with the international developers of the gas cooled fast breeder reactor. As a result of this study, DOE has accepted a Ger.eral Atomic proposal that the upflow concept be adopted for the NASAP-GCFR in preference to the earlier down flow design. In the down flow design fuel elements were cantilevered downward from an upper grid plate with flow forced dowrward through the core under both normal and emergency conditions. Fuel' elements were unrestrained except at the grid plate and were installed or removed from penetra-tions in the lower head of the prestressed concrete reactor vessel (PCRV). The upflow core cooling results in a more conventional reactor design and offers the safety advantage of natural convection core cooling under emergency conditions. The reactor design now provides for fully restrained fuel elements, fuel handling from above, and control rod drive housings similar to HTGRs. The fuel element design is sub-stantially the same as the earlier design including provision for venting fission gases. Some of the other changes are similar to recent changes made to the steam cycle HTGR. Electric motors rather than steam turbines are used to drive the helium circulators, and provisions for reheating steam within the PCRV are eliminated. III EmergencyCoreCoothq Like the HTGR the core auxiliary cooling system (CACS) provides emergency core cooling for both pressurized and depressurized cases. Unlike the HTGR the elevation of the CACS heat exchangers is sufficient that residual heat can also be removed by natural convection. Pony motors on the main circulator shafts provide for an independent, diverse, and redundant means for emergency heat removal under pressurized conditions. In theory, the pony motors could be enlarged to cope with heat removal in the depressurized case, but this is not consid-ered necessary by General Atomic due to the low probability of depressurization occuring simultanecusly with the loss of the CACS. Calculations were presented for natural convection cooling for several cases including malfunctions of primary system valves and 20 percent uncertainty in the decay heat rate. Values of peak fuel rod clad ^'"9 030

l l Steven A. Varga temperature varied from a 1200 F base case to 1441 F for the most conservative case studied. The RATSAM code developed by GA was used for these calculations. RATSAM is being compared against the similar F.R.G. code PHE0 TON and a'so against test data for the CO, cooled Sizewell reactor. Other means for natural convection veriYi-cation used were primary and secondary system simulation tests, data from component tests, and approach to power tests in the demonstra-tion plant. ) I. j:s nn, w Peter M. Wil'liams, Project Manager Division of Project Management '"'9 031

LIST OF ATTENDEES NAME ORGANIZATION Walter C. Lipinski Argonne National Laboratory David R. Buttemer HBA Archie P. Kelley, Jr. HBA Louis Welsnans DOE Colin Fisher GA Brent E. Boyack GA Paul Hunt GA Richard A. Moore GA Aris Papadopenlis NUS Hans Ludewing BNL Peter M. Williams NRC ~"'9 032

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