ML19319A610
| ML19319A610 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/24/1979 |
| From: | Streeter J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Jordan E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML19274F884 | List: |
| References | |
| NUDOCS 7908200172 | |
| Download: ML19319A610 (15) | |
See also: IR 05000346/1978017
Text
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- April 24,1979 ..;r- _ . _ . . I r ::: Doeket No. 50-346 . _ . _ . . m= MEMORANDUM FOR: E. L. Jordan, Assistant Director for Technical Programs, [Jfi' Division of Reactor Operations Inspection, IE 5ll .
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( THRU: / R. F. ishman, Chief, Reactor Operations and Nuclear Ei:[ upport Branch { FROM: J. F. Streeter, Chief, Nuclear Support Section 1 h. [7!== ..:: SUBJECT! INITIAL CORE SELECIIVE LOADING AT DAVIS-BESSE NUCLEAR STATION
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UNIT 1 (A/I F30400H2) f f:m REFERENCES: . (1) Memorandum from J. F. Streeter to R. Woodruff , if .. ..._ dated 7/7/78 El:J[" <(2) Memorandum from E. L. Jordan to J. F. Streeter, hii ~ dated 3/1/79 [" - (3) Memorandum from J. S. Creswell to J. F. Streeter, [M;=..: dated 3/16/79 l - (4) IE Report 50-346/78-17, paragraph 8 g.g (5) IE Report 50-346/79-04, paragraph 2 {{ (6) BrJ-1420, " Davis-Besse Unit 1 - Fuel Densification L: Report", dated 9/75 [{ (7) BWT-1467, " Davis-Besse Unit 1 Selective Fuel Loading", f? dated 2/2/77 h= (8) Memorandum fran P.S. Check to K. Seyfrit, dated 3/28/7/8 !5 le We have reviewed Reference (2) which you sent to us in response to Reference ' ( 1) . As Reference (3) indicates, the inspector who originally identified 7 the potential problem does not believe Reference (2) is an adequate response = _ = .
to Reference (1) and believes the matter should be revie.wed by NRR. The inspector believes NRR needs to give particular attention to determining the effect of Reference (7) on the conclusions NRR reached in Reference (8). . . = Based upon evaluations performed by the lienesee and vendor as described 5E in References (4) and (7) and upon the inspector's telephone communicaticas H with technical personnel in NRR, I do not believe a safety concern of any - consequence exists. However, recognizing the inspector's expertise in the [15l[; core physics area, I request that this natter be forwarded to NRR for rev'ew. c' =
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C ~ 5s E. L. Jordan 2 April 24, 1979 P KE TKh We have been unable to substantiate some GGR values used in Paragraph 2 of
Reference (2) . The average LHGR value of 6.274 W/f: appears to be based __ . = on a design 100% power of 2772 MWe,12 feet of active fuel,177 fuel assem- _.- blies, and 308 fuel rods per assembly (2,772,000 m = 6.274 W/f c). However, - - - (177)(208)(12 feet) the average GGRs given in FSAR Table 4-21 and in Reference (6) are less than that value. If one uses the 6.274 W/f t value and assumes the plant is
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operating at the maximum allowable peaking factor of 2.94 at the start of a 25 short term transient such as dropped rod event, at 112% (design overpower 65 condition) power the maximum LHGR would be 20.66 W/f t. (6.274 W/ft x 5% 1.12 x 2.94 = 20.66 W/f t) . This is of course unacceptable since 20.66 W/ft % is above the fuel melt GGR limit for all fuel in the core and some center y line fuel melting would result. If one uses the 6.143 W/f t value given AE in Reference (6) and the same conditions as before, the maximum LEGR would gjj be 70.23 W/ft and below the fuel melt LHGR limit for all but 2 of the fuel " assemblies in the core. If these 2 assemblies were selectively loaded, as " they have been, in minimum peaking areas of the core as recot: mended in .= . = = . Reference (7), fuel melt GGR limits would not be expected to result under 3[ Condition I or Il events. The determination of the proper average MGR should s: be a result of the NRR review. -~ j5 =E During the NRR evaluation (Reference (8)) of the licensee's evaluation of the fu dropped rod startup test results, NRR concluded that the licensee's technique in extrapolating the LHGR to full power was acceptable and that use of plant p;;;;;; computer values for certain core phys,1cs parameters was acceptable. However, a RIII followup of this extrapolation technique indicated the technique elimi- -gg nated conservatisms without a computer test case having been run to demon- l==:, strate that the computer was capable of either accurately or conservatively h. monitoring the core parameters in question. This finding was documented in /=K Reference (5). The NRR Project Manager is presently having this matter hii. reviewed within NRR. Information on this effort is provided for clarity [? since the inspector has recommended that the NRR conclusions set forth in {_ Reference (8) be reviewed in light of Reference (7).
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For your information, copies of References (3), (4), (5) and (8) are en- Y closed. Sh.--- ,.= J .yF. Streeter, Chief . 2=.;; - Nuclear Support Section 1 M;; bSi% Enclosures: As stated IEE55 [EE cc w/o encl: N.55. R. F. Heishman '
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