ML20202B232

Responds to 960313 Memo Requesting NRR Technical Assistance to Determine Adequacy of NMPC Revised Design Calculations for Reactor Bldg & Turbine Bldg Pressure Relief (or Blowout) Panels.Partially Withheld

ML20202B232
Person / Time
Site:	Nine Mile Point
Issue date:	12/01/1997
From:	Boger B NRC (Affiliation Not Assigned)
To:	Hehl C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20202B115	List: ML20202B112 ML20202B145 ML20202B172 ML20202B232
References
FOIA-97-375 NUDOCS 9712030079
Download: ML20202B232 (11)

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1 MEMORANDUM TO: Charles W. Hehl, Director Division of Reactor Projects, Region i FROM: Bruce E Boger, Director Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation

SUBJECT:

TECHNICAL ASSISTANCE REQUEST REGARDING REACTOR AND TURBINE BUILDING RELIEF PANEL DEFICIENCY, NINE MILE POINT NUCLEAR STATION, UNIT NO.1 (TAC NO. M94858)

By memorandum dated March 13,1996, DRP requested NRR technical assistance to determine the adequacy of Niagara Mohawk Power Corporation's (NMPC) revised design calcula' ions for the Reactor Building and Turbine Building pressure relief (or " blowout") panels.

The cc;culations are related to an event (LER 50-220/95-05, dated November 30,1995) regarding the licensee's discovery, in October 1993, that the panels would not blow out at the design pressure of 45 psf because the bolt fasteners for the panels were larger, and had a higher ultimate strength, than designed. The licensee's initial 1993 engineering calculation of this condition erroneously determined that the Turbine Building panels and Reactor Building panels would blow out at 60 and 53 psi, respectively, to relieve intemal building pressure prior l to structural failure of the buildings, and the panels were declared operable However, during a refueling outage in March 1995, the licensee discovered that an error had been made in 1993 regaroing the design ec&umotion for load distribution The licensee's revised 1995 calculations determined that the relief paneb would not blow out until the intemal building pressuro l

exceeded the minimun documented building structural design of 80 psf. Based on these calculations and before restarting Unit 1 in 1995, the licensee reported the condition to the NRC I and removed every other bolt from the panels to reduce their blowcut point to a value below the l

i documented building structural capability.

! By memorandum dated May 1,1996, Region i supplemented the request for technical l assistance to include two items arising from the related violations of EA 96-079 One item asked whether NMPC was correct in its 10 CFR 50.59 interpretation (apparently based on i

NSAC-125) that safety margins were not reduced because the actual blowout value (if not in j

error, which it was) was still under 80 psf The second item asked whether a 10 CFR 50.59 safety evaluation is needed before maning changes to restore a commitment or safety condition l

' consistent with the original intent of a design. In add, tion, with respect to a public meeting with the licensee on January 6,1997, Region I requested NRR support regarding the reportability of the panels being outside of their design basis.

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9712030079 971201 +

PDR FOIA WILLI APf97-373 PDR ,

6sleausev of Licensee's Calculatigna The licensee's 1993 and 1995 engineering calculations involving the Reactor Building blowout panels and Turbine Building blowout panels have been reviewed by NRR's Mechanical Engineering Branch (EMEB) and Civil Engineering and Geosciences Branch (ECGB). NRR's Containment Systems and Severe Accident Branch (SCSB) provioed technical support to EMEB and ECGB for this review. The pincipal reviewers were David Jeng (ECGB), Mark Hartzman (EMEB), and William Long (SCSB).

Enclosures 1 and 2 are the safety evaluations by ECGB and EMEB, respectively, in these SEs, the staff verified that the licensee's 1993 calculations were technically inadequate. The staff also found tha, the subsequent 1995 calculations, that were intended to correct the deficiencies in the 1993 calculations, were also technically inadequate. After several discussions with the licensee and revised submittals by the licensee to correct technicai deficiencies, the ECGB at:d EMEB reviewers conducted a site audit of the calculations and performed a walk down of the panels. After additional submittals by the licensee resolved issues raised by the NRC during the audit, the NRC staff concluded that the revised calculations were technically adequate.

Using the accepted methodology, the upper bound blowout pressure was determined to be 65 psf fer the Reactor Building panels with every other bolt removed (or 128 psf before the bolts were removed), and 62 psf for the Turbine Building panels after every other bolt was removed (or 122 pr.f before the bolts were removed). The revised ultimate lower bound cr.pacity of the Reactor Building was 117 psf while the corresponding capacity for the Turbine Building was 135 psf. The staff concludes in the SEs that there is ample safety margin for pressure relief of the buildings, based on the current number and size of the bolts.

The above results also show that before every other bol' was removed in 1995, the Reactor Building panels might not have contribuited to the overpressure protection of the Reactor Building (i.e., the panel's mar.imum blowout pressure exceeded the building's lower-bound ultimate capacity), and that the margin of saiety for the Turbine Building during this time was only 13 psf (135 psf minus 122 psf)-significantly less than the 35 psf or more (80 psf or more minus 45 psf) intended by the original design. j 10 CFR 5.0 59 ansi Renortability issues MW. ) Q$ / l SA be.ek At

\ \7/ Gg The NRC's position regarding NMPC's contention that the design bases for intemal building pressure is only to provide pressure relief at or below 80 psf is addressed in Mr. A. Thadani's letter to NMPC dated August XX,1997. As stated in that letter, the NRC considers that the blowout panel pressure of 45 psf is part of the design bases, that 45 psf established the reference for the acceptability of the facility's design, and exceeding 45 psf met the reporting requirements of both 10 CFR 50,72(2)(ii)(B) and 10 CFR 50.73(2)(ii)(B).

Because the NRC's position on the NMP1 violations was not based upon the " margin of safety" test specified in 1C OFR 50.59, the NMP1 citations will not be affected by the resolution of existing differences between industry and NRC as to the pmper definition of this term. By the NRC staff's definition, the margin of safety calculated in 1993 (i.e.,80 psf minus 60 or 53 psf) would be considered significantly less than the design basis margin (80 psf minus 45 psf).

Facility (or procedure) changes by licensees that merely restore a commitment or safety

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condition consistent with the intended design "as described in the safety analysis report" would not noimally require a 10 CFR 50.59 safety evaluation or prior Commission approval. For

. example, NMPC's action in 1995 of removing every other bolt from the panels to restore the blowout pressure to the FSAR specified value of 45 psf did not change the FSAR design description that the NRC staff found acceptable during the operating license review, and did not require a TS change; therefore, it did not require a 10 CFR 50.59 evaluation. The relevant requirements for such restorations are based upon the requirements for the intended design, including as applicable, the quality assurance criteria of 10 CFR 50, Appendix 8.

Enclosure 3 is a bibliogrepy of the principat documents aLociated with this technical review.

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This completes our efforts under TAC M94858 which is now closed.

i Docket No.f 66-220 :

Enclosures:

1. ECGB SE

~ 2. EMEB SE . >

=. 3. Bibliography cc w/ encl:- See next page ,

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DISTRIBUTION:

A. Dromerick T. Harris (e-mail only TLH3)

Docket File PUBLIC S. Little OGC

,oDI 1 R/F D. Hood ACRS S. Varga S West C. Cowgill, Region i G. Bagchi M. Hartzman D.Jeng R. Rothman R. Wessman L. Doerflein, Region i K. Manoly T.Liu E. McKenna M. Satorius DOCUMENT NAME: G:\NMPl\TIA To receive a copy of thic doctment, indicate in the box: "C" - Copy without attachment / enclosure "E" - Copy with attachment /enclor 'e T - No copy orrlu Pn:PDI 1 l Ej LAtPDI-1 l OtPDI 1 l l NAME DMood SLittle ADr eerick

<Date> <Date> <Date> 06/ /97

<Date>

Official Re:ord Copy 1

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! PDI-1 DOCUMENT COVER PAGE l

DOCUMENT NAME: G:\NMPl\TIA ORIGINATOR: D. Hood SECRETARY: R. Laskin

SUBJECT:

TECHNICAL ASSISTANCE REQUEST REGARDING REACTOR AND TURDINE BUILDING RELIEF PANEL DEFICIENCY, NINE MILE P0 INT NUCLEAR STATION UNIT NO. 1 (TAC NO. M94858)

• • • • • ROUTING LIST *****

lBME PAIE

1. / /97

2. D Hood / /97

3. S Little / /97

4. A. Dromerick -

/ /97

5. / /97

6. / /97

7. Secretary - Dispatch PLEASE DO NOT REMOVE W IS SHEET FROM PACKAGE

• CAN THi.S DOCUMENT BE DELETED AFTER DISPATCH 7 YES , NO

1 BIBLIOGRAPHY

1. Final Safety Analysis Report, Nine Mile Point Nuclear Station (Unit 1), dated June 1967, including Section Ill.A.1.2," Turbine Building Design Bases -

Pressure Relief Design," and Section VI.B.1.2," Reactor Building - Design Gasis Pressure Rc!t:f Design."

2. Nuclear Safety Analysis Center report NSAC-125, " Guidelines for 10 CFR 60.59 Safety Evaluations,' datoc June 1989.

3. NRC Generic Letter 91-18 "Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Non-Conforming Conditions and on Operability," dated November 7,1991, including two enclosures from NRC Inspection Manual Chapter 9900, " Technical Guidance."

Enclosure 1 is titled " Resolution of Degraded and Nonconforming Conditions,"

and Enclosure 2 is titled " Operable / Operability: Ensuring the Functional Capability of a System or Component."

4. NUREG-1022, Revision 1 Second Draft,

• Event Reporting Guidelines 10 CFR 50.7' x.150.73," dated February 1994.

5 Updated Final Safety Analysis Re, ort, Nine Mile Point Nuclear Station, Unit 1, dated June 1990, including Section Ill.A.1.2," Turbine Building - Design Bases

- Pressure Ralief Decign," and Section VI.C.1.2, " Reactor Building - Design Bases - Pressure Relief Design."

6. Letter f.om Niagcra Mohawk Power Corporation to U.S. NRC (NMP1L 1007),

dated November 30,1995, forwarding Unit i Licensee Event Report 95-005,

" Building Blowout Panels Outside the Design Basis Because of Construction Error." .

7. Mcr..orandum from Region 1 to NRR dated March 13,1996," Request for Technical Ast! stance on Nine Mile Point 1 Reactor and Turbine Building Blowout Panelt."

8. Letter from U.S. NRC to Niagara Mohawk Power Corporation, dated March 29,1996, forwarding (1) NRC Special Inspection Report Mo. 50-220/96-05; 50 410/96-05. (2) NRC staff questions, and (3)Section V of Enforcement Policy.

9. Memorandem from Region i dated April 2,1996," Notice of Significant Meeting," announcing enforcement conference with Niagara Mohawk Powei Cmoration scheduled for April 12,1996, in King of Prussia, Philadelphia.

i 10. Memorandum from Region 1 to NRR datad May 1,1996, " Request for Technical Assistance on Nine Mile Point 1 Reactor and Turbine Building Blowout Panels Supplement 1."

q 996, " Summary of

11. uemoraneum by D. Hood, U.S. NRC, dated Mm

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Telephone Conversation of May 2,1996, on Reactor and Turbine Building Blowout Panels."

12. Memorandum by D. Hood, U.S. NRC, dated June 7,1996, " Summary of

- Telephone Conversation of May 22,1996, on Reactor and Turbine Building Blowout Panels."

Enclosure 3

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13. Letter from U.S. NRC to Niagara Mohawk Power Corpuration, dated June 18,  !

1996," Notice of Violation and Proposed imposition of Civil Penalty $50,000,"

(EA C6-079).

14. Letter i c m Niagara Mohawk Power Corporation to U.S. NRC (NMPIL 1089),

dated une 26,1996, forwarding Supplement 1 to LER 95-05,

• Building Blowout Panels Outside Design Basis Because of Construction Error."

15. Letter from Niagara Mohawk Power Corporation to U.S. NRC (NMP1L 1096),

dated July 1,1996, Response to Questions in Enclosure .? of Inspection Report 50 220/96-05." 1

16. Letter from Niagara Mohawk Power Corporation to U.S. NRC (NMP1L 1100).

dated July 18,1996, replyirig to June 18,1999 Notice of Violation.

17. Memorandum by D. Hood, U.S. NRC, dated October 7,1996," Trip Report  :

Regarding August 20,1996, Audit of Reactor and Turbine Building Blowout Panel Calculations." t

18. Memorandum by D. Hood, U.S. NRC, dated November 13,1996, " Summary ,

of Telephone Conversation of October 23,1996, on Reactor and Turbine Building Blowout Pan 31s."

19. Letter from Niagara Mohawk Power Corporation to U.S. NRC (NMP1L 1155),

dated November 15,1996, " Response to Trip Report for August 20,1996 Audit of Reactor and Turbine Building Blowout Ptnels."

20. Letter from U.S. NRC to Aiagara Mohawk Power Corporation, dated December 3,1996, " Order imposing a Civil Monetary Penalty - $50,000," (EA 96 079).

21. Memorandum by D. Hood, U.S. NRC, dated January 6,1997," Summary of Telephone Conversation of December 18,1996, on Reactor and Turbine Building Blowout Panels."

22. Letter from Niagara Mohawk Power Corporation to U.S. NRC (NMP1L 1177),

dated January 23,1997," Remittance of Civil, Penalty EA 96-079."

23. Letter frorn U.S. NRC to Niagara Mohawk Power Corporation, dated February 13,1997, 'orwarding a transcript and slides on the January 6,1997, public meeting to discuss issues associated with NRC enforcement action EA ,96-079.

24. Letter from N. Reynolds of Winston and Strawn (Counsel for Niagara Mohawk Power Corporation) to U.S. NRC, dated February 19,1997, requesting -

clarification of reporting requirerr,ents.

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25. NUREG-1606,

• Proposed Regulatory Guidance Related to implementation of 10 CFR 50.59 (Changes, Tests, or Experiments)," published as a draft report for comments April 1997. See e g., Section Ill.S. " Definition of Reduction in Margin of Safety."

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26. Memorandum from R. Zimmerman, U.S. NRC, to ADPR Project Manag( ,

and ADPR Project Directors, dated July 22,1997, " Interim Expectations Rc lated to Oversight of 10 CFR 50.59 Process and FSAR Updates."

27. Letter from A. Thadani, U.S. NRC, to Niagara Mohawk Power Corporation,

'1 dated August XX,1997, responding to request for clarification of reporting 4

requirements.

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