ML20197C477
| ML20197C477 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 01/03/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20197C458 | List: |
| References | |
| FOIA-97-375 NUDOCS 9712240199 | |
| Download: ML20197C477 (16) | |
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' W ETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REACTOR AND TURBlNE PU!LDING BLOWOUT PANELS l NIAGARA M0 HAWK POWER CORPORATION ;
NINE NILE POINT, UNIT 1 i DOCKET No. 50-!!0 t l
t 3.0 RACKGROUnID and the turbine building The at Nine external walls Mlle Point. of 1the Unit reactor (NMP1), containbuilding pane (RB)ls designed to providesure pres (TB) ;
4 relief caused by a nostuinted high energy line break (HELB) within these i buildings.
On November 1, 1995, NMPC submitted Licensee Event Report (LER)95-005 .
'ButidingBlowoutPanelsOutsideoutgnBasisSecauseofConstructionError.' >
(Reference 1 NMPC stated that the relief panel blowout pressures shown in the FSAR were). underestimated and outside the licensin) basis, as a result of an initial construction deficiency and subsequent tecinical deficiencies in design calculations performed in 1993 to correct this deficiency. On the basis of this LER, NRC Region ! conducted a special inspection at NMP3 from -
february 17 through March II 1996, in which several violations of NRC reporting and design control, requirements were idsntified. These were !
detailed in a special inspection report dated March 29, 1996 (Reference 2). A :
predecisional enforcement conference was held on April 12 1996 to discuss these violations. In support of this conference RegionIreque,stedonMarch 13 1996 assistancefromNRRinevaluatingthelechnicaladequacyoftheNMPC calculatons.
2.0 IVALUATION The RB panels are approximately 44' by 19'. The TB panels are approximately i 20' by 20'. Each panel consists of t'-wide by 19' or 20' long horizontal fluted floor decking panel segments, joined and crim>ed to form the complete panel. Each panel In both buildings was assumed to l>e attached to the l building frame at the top and bottom by bolts in shear, and to the butiding '
colurns at the sides by bolts under combined tension and shear.
2.1 1993 CALCULATIONS Two sets of calculations were performed by NMPC in 1993. The purpose of the I first set, dated 8/23/93, was to determine the blowout pressure c:.pacity of the panels based on the number and site (3/16' diameter of the bolts shown ontsede:Igndraaings,andtheminimumultimatetensile)strengthforthebolt material. ,
The first set of 1993 calculations were based on the following assumptions
- Solt minimum ultimate tensile strength;
- Linear bolt tension shear interaction curve; !
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- Panels represented as 'two way' simply supported isotropic plates without in plane membrane action. The in-plane panel edge loads acting on the top and bottom bolts were taken equal to the edge transverse reactions.
The failure mode was detemined to be by shear-tension interaction of the side bolts, and the blowout pressures were determined as 39 psf for the RB and 43 psf for the T8.
The.second Thes6 set datedwere calculations 11/5ba/93, revised the calculations of the first set. i except that the as-installed size of the boltssed1/4' ondi.
the same assumptions,ltimate meter) and bolt u strength values obtained from tests, were use(d. However, because of the changed bolt sizs, the poverning failure mode was deteruined to be shear / tear of the panel sheet meta bolted to the top and bottom bolts. The blowout pressures were determined as 53 psf for the RB and 60 for the TB.
These pressures were lower than the building structural design pressure stated in tne FSAR (80 psf .
the licensing basis) blowout pressures were exceeded, on the basis that theThe pressurization,due to a HELB was not a design basis for NNPl.
2.2 1995 CALCULATIONS NNPC reevaluated the 1993 calculations and operability determination in 1995, .
and identified a number of technical deficiencies in the assumptions and the calculations. The blowout pressures were calculated as 91 psf for the RB and 89 ssf for the TB. To bring the blowout pressures into conformance with the FSAt values, every other bolt in the side connections was removed, which halved these values. -
The 1995 blowout calculations were based on the actual bolt size and ultimate strength, and the following assumptions:
- Linear bolt tension-shear interaction curve; Panel segments represented as "one way' simply supported beams, each considered to act separately from adjacent segments.
2.3 IVALUATION OF ASSU@ TIONS The staff reviewed the assumptions on whleh the analyses were based, and concluded that they did not account for realistic deformation bahavior of t >
bolts and the panels, and that the actual blowout pressures of the parels were therefore underestimated and not an upper bound.
The representation of the panels as simply supported 'two way' isotropic plates is technically deficient since it does not correspond to the as-built type shear connection at the top and bottom of the panels and the difference in horizontal and vertical stiffness. The panels develop longitudinal membrane action which exists as long as the horizontal panel segments remain joined and crimped. In addition the panels are considerael)stifferinthehorizontaldirectionthanInthevertical l
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3 i direction. The panels should therefore have been analyzed as shply supported 'two way' orthotropic plates reflecting finite membrane action in the vertical direction. l The representation'of the panti b 'one we simply su beams is valid only if the crimped joints b* tween pane $ ported segments of thehorizontali onthe{onewa'two wa ' panels open up at a lower pressure than that determined!
, calculations. y' assumption. This pressure was not determined in these :
The use of a linear tension / shear interaction curve to represent bolt behavior is not conservative in this case, since the actual bolt -
strengths are underestimated. The bolt behavior in the calculations i whould have been represented by an elliptic tension / shear interaction '
cutve.
A number of ethar technical deficiencies were identified, and NMPC was ;
requested te Wrovide a more realistic estimate of the blowout pressures, j t,1 N $tD D T CONFERfNtt A predsel8%hal enforcement conference was Md on April 12, 1996, at Region 1, at which the technical deficiencies were discussed with NMPC. The staff expretsed its concerns regarding the assumptions on which the calculations !
were based, and NMPC was requested to provide a revised assesement of the blowout pressures for both buildings, based on realistic assumptions.
At this conference, the Region ! staff also e.xpressed a concern regarding the safety of the storspe condensate tanks housed in t% auxiliary turbine building. This but ding is located at a lower elevation than that of the blowout pant)4, and could potentially be subject to impact and piercing of the roof by the blownout panels. NMPC comited to perform a safety evaluation of this event.
2.5 AUDIT AT WINE MILE PolNT. UNIT ONE On August 20, 1996 the NRA staff perfomed an audit at NMP3 of the revised NMPC calculations., The staff reviewed the calculation of the blowout pressure of the panels for both bul' dings. These calculations were performed based on the following assumptions: t The bolt behavior is governed by an elliptic tension / shear interaction equation.
The b>1t ultimate strength was taken as the highest determined from the [
1993 tests.
The panel segments are considered as 'one way' simply supported beams.
The staff reviewed the justification for the last assumption. NMPC perforu d calculations showing that the pressure at which the ultimate load capacity of '
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the crimped joints between the panel segments is exceeded is lower than the gressuredeterminedtofalltheboltsonthesideconnectIonsbasedonthe one way' assumstion. The staff found this acceptable, based on its own assessment of t se ultimate load capacity of the crimped joints.
On this basis the panel upper bound blowout pressure for the R8 was deterv.ined as,65 psf, and for the T8 as 62 psf. Based on the initial number of bolts, the panel blowout pressure of the RB was also calculated as 128 ssf i and.for the T8 as 122 psf. The staff found these values in accordance wit) its own assessment of the pressures under similar assumptions as those invoked by NMPC. -
NNPC also recalculated the ultimate lower bound capacities of the buildings as 117 psf for the R8 and 135 psf for the TB. This shows that there is ample safety, margin for pressure relief of the buildings, based on the current number and size of the bolts.
l, 2.6 IMPACT DN THE TURBINE BUILDING ROOF The staff also reviewed the response by NMPC to the concern regardin of a blowout pahel on the TB roof over the condensate storage tanks.g The impact initial response was unacceptable due to a technical error in the analysis.
NMPC corrected this error and prov,ided a revised response, based on a proper Y
analytical evaluation. The results show that the safety of the condensate storage tanks is not compromised by this event since the roof is capable of absorbing plastic deformation. the impact energy of a panel without rupturing or experiencing large and found them acceptable.
The staff reviewed the revised response and the results,
3.0 CONCLUSION
The staff finds that the initial NMPC a proach for calculating the blowout panel pressures was based on technicall deficient simplif.ying assumptions which did not reflect the as-built pane construction and boundary conditions, the mode of load transmittal from the panels to the bolts, and the real behavior under combined loading of the bolts.
- NNPC has provided revised calculations of the upper bound blowout panel pressures.
NMPC has also provided an evaluation of panel impact on the roof of the turbine building. The staff has evaluated the bases of these calculations and finds them reasonable and in accord with current engineering practice.
Principal Contributor: M. Hartaman, NRR/EMEB 415-2755
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[ AEFERENCES l
- 1. Letter of November 30, 1995, from N. L. Rademacher, Nia Ccrporation (NMPC), to the USNRC Document Controlwith Desk,gara enclosedMohawk Pow Licensee Event Report LER 95-05 'Nine Mile Point, Unit 1, Building Blowout Panels outside(the) Design Basis Because of Construction E
- 2. Letter of March 29, 1996, frota R. W. Cooper, !!, DRP, Region I, to B. R.
Sylvia, NMPC, with enclosed NRC Special Inspection Report No. 50-220/96; 50-410/96 05.
- 3. Calculations of the Pressure Capacity of Possure Relief Panels in the Reactor and Turbine Buildings at Nine Mile Point Nuclear Station, Unit 1 i
l (NMP1), dated August 23, 1993, and March 29, 1995.
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i MEMOR/sNDUM 'iO: Charles W. Hehl. Director Division of Reactor Projects. Region 1 FROM: Alexander W. Dromerick, Acting Director Project Directorate 1 1 Division of Reactor Projects 1/11 Office of Nuclear Reactor Regulation
SUBJECT:
TECHNICAL ASSISTANCE REQUEST REGARDING REACTOR AND TURB!NE BUILO!NG REL EF PANEL DEflCIENCY, NINE MILE POINT NUCtLAR STATION, UNIT NO. 1 (NMP1) (TAC NO. M94858)
By memorandum dated March 13, 1996, Division of Reactor Projects (ORP) requested NRR technical assistance to determine the adequacy of Niagara Mohawk Power Corporation's (NMPC) revised design calculations for the Reactor Building and Turbino Building pressure relief f (or " blowout") panels. The cairulatians 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 pounds per square foot (psf) because the bolt fasteners for the panels were larger, and had a higher ultimate strength, than designed. The lic,ensee'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 psf, respectively, to relieve internal building pressure prior to structural failure of the buildings, and the panels were declared operable. However.-during a refueling outage in March 1995. the licen re discovered that an error had been made in 1993 regarding the design assumption for load distribution. The licensee's revised 1995 calculations determined that the relief panels would not blow out until the internal building pressure exceeded the minimum documented building Jtructural design of 80 psf. Based on these calculations and before restarting Unit 1 in 1996, the licensee reported the condition to the NRC and removed every other bolt from the panels to reduce their blowout point to a value below the documented building structural capability.
By memorandum dated May 1,1996. Region I supplemented the request for technical assistance to include two items arising from the related violations of EA 96 079. One item asked *f whether NMPC was correct in its 10 CFR 50.59 interpretation (apparently based on NSAC 125) that safety margins were not reduced because the actual blowout value (if not in error, which it was) was still under 80 psf. The second item asked whether a 10 CFR 50.59 safety /
evaluation (SE) 1s needed before making changes to restere a comitment or safety condition consistent with the original intent of a design. In addition, with respect to a public .
meeting with the licensee on January 6. I M . Recion I requested NRR support regarding the /
reportability of the panels being outside of their design basis.
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2 Adewacy of ticensee's Calculations ,
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 Brarch (EMEB) and Civil Engineering and Geosciences Branch (EC6B). NRR's Containment Systems and Severe Accident Branch (SCSB) provided technical support to EMEB and ECGB for this review. The principal reviewers were David Jeng (ECGB). Mark Hartzman (EMEB).
and William Long (SCSB). ,
Attachments 1 and 2 are tie SEs by ECGB and EMLB. respectively. !n these SEs. the staff verified that the licenses 's 1993 calculations were technically inadequate. The staff also found that 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 technical deficiencies, the ECGB and 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 detemined to be 65 psf for the Reactor Building panels with every other bolt removed (or 12B psf before the bolts were removed), and 62 psf for the Turbine Building panels after every other bolt was removed (or 122 psf before the bolts were removed), The revised ultimate lower bound capacity 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 butidings, based on the current number and site of the bolts.
The above results also show that before every other bolt was removed in 1995, the Reactor Building panels might not have contributed to the overpressure protection of the Reactor Building (i.e., the panel's maximum blowout pressure exceeded the building's lowe 90und ultimate capacity), and that the margin of safety 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, 10 UR 50 59 and Reportability Issues The NRC's position regarding NMPC's contention that the design bases for internal building pressure is only to provide pressure relief at or below 80 psf is addressed in Mr. A.
Thadant's letter to N @C dated September 12, 1997. As stated in that letter, the NRC considers that the blowout Danel pressure of 45 psf is part of the design bases, that 45 psf established the reference foe 3
the acceptability of the facility's design, and exceeding 45 psf met the reporting requirements of both 10 CFR 50.72(b)(1)(11)(B) and 10 CFR 50.73(a)(2)(ii)(B)
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4 Because the NRC's position on the NMP1 violations was not based upon the ' margin of safety * ,
test specified in 10 CFR 50.59. the NMP1 citations will not be affected by the resolution of existing differences between industry and NRC as to the proper definition of this term. By the NRC stcff's definition, the margin of safety calculated in 1993 (1.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 comitment or safety condition consistent with the intended design 'as described in the safety analysis report" would not normally require a 10 CFR 50.59 SE or prior Comission approval. For example, 'j NMPC's action in 1995 of removing every other bolt from the panels to restore the blowout pressure to the final safety analysis report (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 e TS change; therefore, it did not require a 10 CFR 50.59 v ,
evaluation. The relevant requirements for such restorations are based upon the requirements for the intended design, incluutng as applicable the quality assurance criteria of 10 CFR Part 50. Appendix B. l l
Attachment 3 is a bibliography of the principal documents associated with this technical review. ,
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this completes our efforts under TAC N0. M94858 which is now closed.
Docket No, 50 220 Attachments: 1. ECGB SE
- 2. EMEB SE
- 3. Bibliography
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This completes our efforts under IAC NO. M94858 which is now closed. 1 i
Docket No. 50 220 Attachments: 1. ECGB SE
- 2. EMEB SE
- 3. Bibliography I
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l DISTRIBUT10N:
Docket file G. Bagchi PDl 1 R/f R. Rothman J. Lieberman K. Manoly B. Boger M. Satorius E. Adensam f. Akstulewicz J. Zwolinski M. Hartzman A, Dromerick R. Wessman M. Boyle (email only) C. Cowgill. R1 M. Oprendek, Region I L. Doerflein. RI S. Little D. Jeng T. Liu E. McKenna D. Hood B. Norris DOCUMENT NAME: G:\NMPl\M94858.TI A To receive a copy of this document, indicate in the box: "C" - Copy without attachment / enclosure *E" - Copy with attachment / enclosure "N" - No copy OfflCE PMIPDt.) lE LA PDl.1 l D PDial l DD DRPE l PGE B hAME DMood $little ADromerick J2wolinski FAkstulewic OA11 <Date> <Date> <Date> <Date> 09/ /97 Official Record Copy
PDI-1 DOCUMENT COVER PAGE DOCUMENT NAME: G:\NMPl\M94858.TIA ORIGINATOR: D. Hood .
SECRETARY: R. Laskin
SUBJECT:
TECHNICAL ASSISTANCE REQUEST REGARDING REACTOR AND TURBINE BUILDING RELIEF PANEL DEFICIENCY. N!NE MILE POINT NUCLEAR STATION UNIT NO. 1 (TAC NC. M94858)
- ***** ROUTING LIST *****
LE DAII
- 1. D Hood / /97
- 2. S Little / /97
- 3. A Dromerick / /97
- 4. J. Zwolinski / /97 li
- 5. / /97
- 6. / /97
- 7. / /97
- 8. Secretary - Dispatch PLEASE 00 NOT REMOVE THIS SHEET FROM PACKAGE
- CAN THIS DOCUMENT BE DELETED AFTER DISPATCH 7 YES NO
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. BIBLIOGRAPHY !
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- 1. Final Safety Analysis Report. Nine Mile Point Nuclear Station (Unit 1),
dated Jurie 1967, including Section Ill.A.1.2, " Turbine Building -
Design Bases Pressure Relief Design," and Section VI.B.I.2, :
- Reactor Building - Design Basis Pressure Relief Design."
- 2. Nuclear Safety Analysis Center report NSAC 125. " Guidelines for 10 CFR 50.59 Safety Evaluations," dated June 19B9.
- 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 1s citled
" Operable / Operability: Ensuring the functional Capability of a System or Component."
- 4. NURlG 1022. Revision 1 Second Draft " Event Reporting Guidelines 10 CFR 50.72 and 50.73 " dated February 1994.
- 5. Updated Final Safety Analysis Report. Nine Mile Point Nuclear Statior.,
Unit 1, dated June 1996, including Section ll!.A.1.2, " Turbine Building Design Bases Pressure Relief Design," and Section VI.C.1.2, " Reactor Building Design Bases - Pressure Relief Design."
- 6. Letter from Niagara Mohawk Power Corporation to U.S. NRC (NMPIL 1007),
dated November 30, 1995, forwarding Unit 1 Licensee Event Report 95 005, " Building Blowout Panels Outside the Design Basis Because of Construction Error."
- 7. Memorandum from Region 1 to NRR dated March 13. 1996. " Request for Technical Assistance on Nine Mile Point 1 Reactor and Turbine Building Blowout Panels."
B. Letter from U.S. NRC to Niagara Mohawk Power Corporation. dated March 29, 1996, forwarding (1) NRC Special Inspection Report No.
50 220/96 05: 50 410/96 05, (2) NRC staff questions, and (3)Section V of Enforcement Policy.
9, Memorandum from Region 1 dated April 2. 1996, " Notice of Significant ;
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Meeting," announcing enforcement conference with Niagara Mohawk Power Corporation scheduled for April 12, 1996, in King of Prussia, Philadelphia.
- 10. Memorandum from Region 1 to NRR dated May 1, 1996. " Request for Technical Assistance on Nine Mile Point 1 Reactor and Turbine
-Building Blowout Panels - Supplement 1."
- 11. Memorandum by D. Hood. U.S. NRC, dated May 8. 1996. " Summary of Telephone Conversation of May 2. 1996, on Reactor and Turbine Building Blowout Panels."
Attachment 3
- 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."
- 13. Letter from U.S. NDC to Niagara Mohawk Power Corporation, dated June 18.
1996. " Notice of Violation and Proposed Imposition of Civil Penalty -
$50,000." (EA 96 079,.
- 14. Letter from Niagara Mohawk Power Cerporation to U.S. NRC (NMPIL 1089),
dated June 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 (NMPIL 1096).
dated July 3. 1996. " Response to Questions in Enclosure 2 of Inspection Repert 50-220/96 05."
- 16. Letter from Niagara Mohawk Power Corporation to U.S. NRC (NMPIL 1100),
dated July 16, 1996. replying to June 18. 1996 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."
1B. 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 Panels."
- 19. Letter from Niagara Mohawk Power Corporation to U.S. NRC (NMPIL 1155).
dated November 15. 1996. " Response to Trip Report for August 20. 1996 Audit of Reactor and Turbine Building Blowout Panels."
- 20. Letter from U.S. HRC to Niagara Mohawk Power Corporation dated Decemoer 3. 1996. " Order Imposing a Civil Monetary Penalty -
$50.000 " (El. 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
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- Building Bleacut Panels."
- 22. Letter from Niagara Mohawk Power Corporation to U.S. NRC (NMPIL 1177),
dated January 23. 1997. " Remittance of Civil Penalty EA 96 079."
- 23. Letter from U.S. NRC to Niagara Mohawk Power Corporation, dated Feoruary 13. 1997. forwarding a tra, script and slides on the
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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 requirements.
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. 3 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 III.S. " Definition of Reduction-in Margin of Safety."
26, Letter from Niagara Mohawt Power Corporation to U.S. NRC endorsing -
comments of July 7,1997, by Nuclear Energy Institute and Winston &
Strawn on NUREG 1606.
- 27. Memorandum from R. Zimmerman, U.S. NRC, to ADPR Project Managers and ADPR Project Directors. dated July 22,1997, " Interim Expectations Related to Oversight of 10 CFR 50.59 Process and FSAR Updates."
- 28. Letter from A. Thadant, U.S. NRC, to Niagara Mohawk Power Corporation, dated September 12, 1997. responding to request for clarification of reporting requirements.