NUREG/CR-1018, Response in Opposition to Jf Doherty Contention 45 Re Need for Addl Lateral Support in Reactor Core.Intervenor Misread NUREG/CR-1018 by Relying on Analysis & Recommendations Relating Solely to Pwrs.Certificate of Svc Encl

Response in Opposition to Jf Doherty Contention 45 Re Need for Addl Lateral Support in Reactor Core.Intervenor Misread NUREG/CR-1018 by Relying on Analysis & Recommendations Relating Solely to Pwrs.Certificate of Svc Encl

ML19256G350
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	12/20/1979
From:	Sohinki S NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
To:	Atomic Safety and Licensing Board Panel
References
NUDOCS 7912310135
Download: ML19256G350 (4)

Text

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0/79 UNITED STATES OF AfiERICA NUCLEAR REGULATORY C0fiMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of HOUSTON LIGHTING & POWER COMPANY

)

Docket No. 50-466

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(Allens Creek Nuclear Generating

)

Station, Unit 1)

)

NRC STAFF'S RESPONSE TO JOHN F. DOHERTY'S CONTENTION #45, NEED FOR ADDITIONAL LATERAL SUPPORT IN THE REACTOR CORF The NRC Staff opposes what is, in effect, the December 3,1979 motion to admit a new contention (#45) filed by John F. Doherty in the captioned pro-ceeding. The newly proffered contention concerns an alleged insufficiency in the lateral support of the Allens Creek reactor core resulting in an inability of the core to withstand the lateral force due to flashing "which occurs near the end of the subcooled blowdown portion of the LOCA transient."

The contention is based upon Mr. Doherty's review of NUREG/CR-1018 and the recomendations contained therein.

However, Mr. Doherty has, we believe, misread the cited report.

In using NUREG/

CR-1018 as the basis for his contention, he has relied upon the analysis and recommendations specified for a PWR (Allens Creek is, of course, a BWR).

Appendix A of the report (p. A-2), in the section " Crossflow Velocity Force Calculation,"specifically indicates that:

Differential flashing may cause lateral fluid velocities in the core region of a pressurized water reactor during a postulated loss-of-coolant accident near the end of the subcooled blowdown region.

l 7@l0 t =s'=5

This calculation is used to determine the additional margin for lateral loads as stated:

(a) 30% for an SSE impact load or (b) a factor of 1.3 for the LOCA or combined SSE-LOCA analyses. The report indicates (p.14) that:

Additional margin, such as discussed above for lateral LOCA forces is not applicable if a linear bounding analysis can be justified.

Linear response analysis techniques are well defined.

For the Allens Creek facility, such a linear analysis has been approved by the Staff after its review of the GE topical report NEDE-21175-P, "BWR 6 Fuel Assembly Evaluation of Combined SSE and LOCA loadings."

In the Staff's evaluation of that report (see attached letter from 0.D. Parr to G.G. Sherwood, dated May 17, 1978), we state:

Unlike in a PWR, the blowdown load due to pipe rupture in a BWR is expected to be small. Consequently, the analysis revealed no lateral impacting between the fuel assemblies.

This makes the fuel assembly response independent of gap size between the assemblies and permits a relatively simple linear analysis.

Since the analysis and recommendations on which Mr. Doherty bases his contention relate solely to PWRs, the Staf f urges the Board to deny the motion to admit contention #45.

Respectfully submitted,

~k k-

[} L.

l l Stephen M. Schinki Counsel for NRC Staff Dated at Bethesda, Maryland, this 20th day of December,1979.

UNITED STATES OF AMERICA NUCLEAR REGULATURY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of HOUSTON LIGHTING & POWER COMPANY Docket No. 50-466 (Allens Creek Nuclear Generating

)

Station, Unit 1)

)

CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S RESPONSE TO JOHN F. D0HERTY'S CONTENTION #45, NEED FOR ADDITIONAL LATERAL SUPPORT IN THE REACTOR CORE" in the above-captioned proceeding have been served on th'e following by deposit in the United States mail, first class, or, as indicated by an asterisk by deposit in the Nuclear Regulatory Commission internal mail system, this 20th day of December,1979:

Sheldon J. Wolfe, Esq., Chairman

• Richard Lowerre, Esq.

Atomic Safety and Licensing Board Panel Asst. Attorney General for the U.S. Nuclear Regulatory Commission State of Texas Washington, DC 20555 P.O. Box 12548 Capitol Station Dr. E. Leonard Cheatum Austin, Texas 78711 Route 3, Box 350A Watkinsville, Georgia 30677 Hon. Jerry Sliva, Mayor City of Wallis, Texas 77485 Mr. Gustave A. Linenberger

• Atomic Safety and Licensing Board Panel Hon. John R. Mikeska U.S. Nuclear Regulatory Commission Austin County Judge Washington, DC 20555 P.O. Box 310 Bellville, Texas 77418 R. Gordon Gooch, Esq.

Baker & Botts Mr. John F. Doherty 1701 Pennsylvania Avenue, N.W.

4327 Alconbury Street Washington, DC 20U06 Houston, Texas 77021 J. Gregory Copeland, Esq.

Mr. and Mrs. Robert S. Framson Baker & Botts 4822 Waynesboro Drive One Shell Plaza Houston, Texas 77035 Houston, Texas 77002 Mr. F. H. Potthoff, III Jack Newman, Esq.

1814 Pine Village Lowenstein, Reis, Newman & Axelrad Houston, Texas 77080 1025 Connecticut Avenue, N.W.

Washington, DC 20037 D. Marrack 420 Mulberry Lane Carro Hinderstein Bellaire, Texas 77401 8739 Link Terrace Houston, Texas 77025 1661 280

=e-o e--

Margaret Bishop Texas Public Interest 11418 Oak Spring Research Group, Inc.

Houston, Texas 77043 c/o James Scott, Jr., Esq.

8302 Albacore Glen Van Slyke Hot!ston, Texas 77074 1739 Marsnall Houston, Texas 77098 Brenda A. McCorkle 6140 Darnell J. Morgan Bishop Houston, Texas 770/4 a

11418 Oak Spring Houston, Texas 77043 Mr. Wayne Rentfro P.O. Box 1335 Stephen.A. Doggett, Esq.

Rosenberg, Texas 77471 Pollan, Nicholson & Doggett P.O. Box 592 Rosemary N. Lemmer Rosenberg, Texas 77471 11423 Oak Spring Houston, Texas 77043 Bryan L. Baker Charles Andrew Perez

,1118 Montrose Houston, Texas 77019 1014 Montrose Blvd.

Houston, Texas 77019 Robin Griffith 1034 Sally Ann Leotis Johnston Rosenberg, Texas 77471 1407 Scenic Ridge Houston, Texas 77043 Elinore P. Cummings 926 Horace Mann Atomic Safety and Licensing

• Rosenberg, Texas 77471 Appeal Board U.S. Nuclear Regulatory Comnission Mrs. Connie Wilson Washington, DC 20555 11427 Oak Spring Houston, Texas 77043 Atomic Safety and Licensing

• Board Panel Patricia L. Streilein U.S. Nuclear Regulatory Commission Route 2, Box 398-C Pashington, DC 20555 Richmon, Texas 77469 Docketing and Service Section

• Carolina Conn Office of the Secretary 1414 Scenic Ridge U.S. Nuclear Regulatory Co rnission Houston, Texas 77043 Washington, DC 20555 Mr. Robert Alexander Mr. William J. Schuessler 10925 Briar Forest #1056 5810 Darnell Houston, TX 77042 Houston, Texas 77074 b

/

w Stepherf M. Schinki Counsel for NRC Staff 1661 281

.a Distribution w/ enclosure -

Central File u M. Rushbrook S.Kim NRC PDR P.. Capra LWR #3 File R. Boyd BCC:

JBuchanan D. Crutchfield TAbernathy R. Diggs ACRS (16)

S. Hanauer Central Files - Topical Reports Dr. G. G. Shemood Manaaer - Safety and Licensing General Electric Company 175 Curtner Avenue San Jose, California 95114

SUBJECT:

STAFF EVALUATION OF TOPICAL REPORT tlEDE-21175-P Al:D !!ED0-21175 We have completed our review, through Amendment 2, of General Electric topical report llEDE-21175-P (proprietary) and llEDD-21175 (non-proprietary version), " BUR /6 Fuel Assembly Evaluation of Conbined SSE and LOCA Loadings." This report describes: (1) seismic and blowdown loadings, (2) analytical methods used to determine fuel assembly structural response to the loads, and (3) design limits used to detennine the a4quacy of BUR /6 fuel assemblies.

Based on our review, we conclude that the analytical methods described in this report are acceptable for reference in license applications as discussed in the enclosed staff evaluation. We have not reviewed the generic fuel assembly design limits since we have not yet completed developing general acceptance criteria for the design limits. The seismic and blowdown loads were not reviewed and must be considered on a plant-by-plant basis.

In addition, the analysis of the potential fuel bundle liftoff from the core plate for a postulated steam line break will be evaluated seaarately and a report will be issued on the subject at a later time.

The staff does not intend to repeat its review of this report when it is referenced in specific license applications, except to assure that the report is applicable to the specific plant involved. When the proprietary report is used as a reference, both the proprietary and the non-proprietary version of the report must be referenced.

In accordance with established procedure, it is requested that General Electric issue a revised version of this report to include: Amendments 1 and 2, an supplementary information provided acceptance letter, and the staf f e DUPLICATE DOCUMENT

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