ML20138G181
| ML20138G181 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/25/1997 |
| From: | Sumner H SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR HL-5377, TAC-M69451, TAC-M69452, NUDOCS 9705060198 | |
| Download: ML20138G181 (20) | |
Text
^
I Lewis Sumner Sruthern Nntear
. . Vice President Operating Company,Inc.
,
- Hatch Project Support 40 Invemess Parkway Post Othee Box 1295 Binningham, Alabama 35201 Tel 205992 7279 Fax 205 992.0341 SOUTHERN aa I COMPANY April 25,1997 EscrgyroScrre nurWld" Docket Nos. 50-321 50-366 liL-5377 Tac Nos. M69451 i M69452 l l
U. S. Nuclear Regulatory Commission !
ATTN: Document Control Desk Washington, D. C. 20555 Edwin I.11atch Nuclear Plant Response to Supplemental Request for Additional Information on the Resolution of Unresolved Safety issue A-4fz Gentlemen:
By letter dated January 30,1997, the Nuclear Regulatory Commission (NRC) staff requested additional information regarding the resolution of Unresolved Safety Issue (USI) A-46. By letter dated March 4,1997, the response was requested to be extended an additional 30 days in order to solicit industry review of the Southern Nuclear Op: rating Company (SNC) l response. The information requested is provided in the enclos. ire and attachment. I Sincerely, ,
1
&wYJ \
- 11. L. Sumner, Jr.
DLM/ld
Enclosure:
Unresolved Safety Issue A-46 l
l cc: Southern Nuclear Oneratine Comnanv i Mr. P.11. Wells, Nuclear Plant General Manager f
NORMS
_Acg{
US Nuclear Remdatorv Commission. Washineton. D.C.
Mr. K. Jabbour, Licensing Project Manager - liatch US Nuclear Remdatorv Commission. Reelon H lli11 mm llll llIllllll o < m .
ll Mr. L. A. Reyes, Regional Adminisaator OGO007 Mr. B. L. liolbrook, Senior Resident inspector - Hatch 9705060198 970425 PDR ADOCK 05000321 P PDR ,
W Enclosure Response to Supplemental Request for Additional Information on the Resolution of Unresolved Safety issue A-46
- 1. NRC Request With regard to the concern involving the use of proper seismic demand for equipment within '
40 feet above the effective grade level, the staff's assessment is that your response of August 23,1996, did not address the requested information. In the SQUG/NRC meeting held on August 28,1996, the staff elaborated its concern and the primary focus of the request for additional information (RAI) questions. As a result of considerable discussion on the subject, the staff agreed to clarify the question. The following represents the revised question, which is being forwarded to the affected USI A-46 licensees for their response:
Referring to the in-structure response spectra provided in your 120-day response to the NRC's request in Supplement No. I to Generic Letter (GL) 87-02, dated May 22,1992, the following information is requested:
- a. Identify structure (s) that have in-structure response spectra (5% critical damping) for elevations within 40 feet above the effective grade, which are higher in amplitude than 1.5 times the SQUG Bounding Spectrum.
- b. With respect to the comparison of equipment seismic capacity and seismic demand, l indicate which method in Table 4-1 of the Generic Implementation Procedure (GIP) j GIP-2 was used to evaluate the seismic adequacy for equipment installed on the l corresponding floors in the structure (s) identified in Item (a) above. If you have elected I to use method A in Table 4-1 of the GIP-2, provide a technicaljustification for not using the in-structure response spectra provided in your 120-day response. It appears that some A-46 licensees are making an incorrect comparison between their plant's safe shutdown carthquake (SSE) ground motion response spectrum and the SQUG Bounding Spectrum. The SSE ground motion response spectrum for most nuclear power plants is defined at the plant foundation level. The SQUG Bounding Spectrum is defined at the free field ground surface. For plants founded on deep soil or rock, there may not be a significant difference between the ground motion amplitudes at the foundation level and those at the ground surface. However, foi sites where a structure ,
is founded on shallow soil, the amplification of the ground motion from the foundation level to the ground surface may be significant,
- c. For the structure (s) identified in item (a) above, provide the in-structure response spectra designated according to the height above the effective grade. If the in-structure IIL-5377 E-1
l Enclosure Unresolved Safety Issue A-46 1
l response spectra identified in the 120-day response to Supplement No. I to Genenc j Letter 87-02 was not used, provide the response spectra that were actually used to i verify the seismic adequacy of equipment within the structures identified in item (a) l above. Also, provide a comparison of these spectra to 1.5 times the Bounding i Spectrum.
SNC Resnonse There are no Seismic Category 1 structures at the Edwin I. Hatch Plant that have in-structure response spectra (IRS) (5% critical damping) for elevations within 40 feet above the effective grade, which are higher in amplitude than 1.5 times the Seismic Qualification )
Utility Group (SQUG) Bounding Spectrum. By letter dated September 16,1992, a response j to Generic Letter 87-02, Supplement No.1, stated that when using IRS for resolution of USl A-46, the Plant Hatch seismic margin earti quake IRS times one-half will be used. All of these IRS for resolution of USl A-46 up to within 40 feet above the effective grade have spectral acceleration amplitudes below that of spectral acceleration amplitudes of 1.5 times the SQUG Bounding Spectrum.
Since there are no seismic category structures with IRS which are higher in amplitude than 1.5 times the SQUG Bounding Spectrum and the IRS identified in the 120-day response to Generic Letter 87-02, Supplement I was used to evaluate seismic adequacy, a response to requests Ib and Ic is not applicable.
- 2. NRC Request In reference to your letter dated July 31,1996, regarding the completion of actions in l accordance with Supplement I to GL 87-02, you stated that you have changed the plant licensing bases prior to receipt of the plant-specific safety evaluation. The changes performed were related to the Final Safety Analysis Report (FSAR), for Units 1 and 2, that l incorporate the USl A-46 GIP methodology for verifying the seismic adequacy of new, replacement, and existing electrical and mechanical equipment. We request that you submit, for the staft's review, the complete documentation associated with your evaluation of the unreviewed safety question associated with 10 CFR 50.59 for carrying out the FSAR changes for equipment qualification.
SNC Resnonse 1
The requested documentation associated with SNC's 10 CFR 50.59 evaluation for changing the FSAR to incorporate the GIP methodology for verifying the seismic adequacy of new, replacement, and existing electrical and mechanical equipment is provided as an attachment to this enclosure.
HL-5377 E-2
i Enclosure Unresolved Safety Issue A-46 It is SNC's position that the GIP was approved by the NRC Staffin Supplemental Safety )
Evaluation Report No. 2, dated May 22,1992, as an acceptable method of demonstrating the seismic adequacy of equipment within its scope. The GIP methodology differs from the methodology used in original plant licensing during the 1970s in substantial and j fundamental respects. Any such comparison of the two methodologies must be made at the program level to evaluate compliance with appropriate regulations concerning seismic adequacy. Additionally, SNC's conclusions that the subject FSAR change does not represent an unreviewed safety question reflect the findings of previous site-specific NRC ,
Staf f reviews associated with seismic design issues and the seismic margins program. l l
- 3. NRC Request )
)
Your response to NRC staff's question No. 9, transmitted by our letter dated June 27,1996, stated in part that: "For Plant Hatch, the only issue identified is the possible loss of the normal lighting system, in which case the operators rely upon emergency lighting or hand- )
held lights to perform their duties."
Describe what, if any, other barriers to successful operator performance to reset the DG relays were considered and resolved as part of the seismic and relay evaluation. In addition to lighting, discuss what, if any, other hazards or environmental factors such as temperature, ,
humidity, debris, or damaged structures, which could inhibit an operator from accomplishing the task of resetting the DG relays in the time-frame allotted, were considered. !
SNC Resnonse The potential for barriers such as damaged equipment or structures which could inhibit an operator's ability to access plant equipment was considered during the development of the SQUG GIP and found to be very unlikely. Earthquake experience has shown that typical industrial grade equipment and structures are inherently rugged and are not susceptible to damage at A-46 plant SSE levels. It is considered very unlikely that operators will be faced with hazardous or unfamiliar circumstances which are not covered by existing plant procedures and training. It is for this reason that the GIP, in Section 3.2.7, allows operator action as a means of achieving and maintaining a safe shutdown condition provided procedures are available and the operators are trained in their use.
The Diesel Generator (DG) building is a Seismic Category I structure. As stated in the Edwin 1. Hatch Plant Individual Plant Examination of External Events (IPEEE) Report, transmitted to the NRC by letter dated January 26,1996, the Seismic Margin Assessment (SMA) demonstrated that the Hatch DG building has a High-Confidence-Low-Probability-of-Failure (HCLPF) level of at least 0.3 g peak ground acceleration. Therefore, it has been demonstrated that the DG building will remain intact with no structural damage that could hinder operator actions to reach the DG relays. The DG building is not physically attached HL-5377 E-3
Enclosure Unresolved Safety Issue A-46 l
to any other structure. There are no other structures or equipment in the area that are expected to block access to the main DG building entrance. In the postulated event that the main entrance located on the south side of the building is not accessible, there are six additional entrance doors located on the west side of the DG building. These doors open directly into each of the DG switchgear rooms where the relays in question are located.
All equipment and structures inside the DG building are designated as either Seismic Category I or II/I, which assures that they will be prevented from falling or moving in such a way that they would hinder movement of the plant operators. Additionally, the USI A-46 and IPEEE walkdowns did not reveal any stored items inside the DG building that would hinder the movement of the operators. This is especially true in the DG and switchgear rooms. These rooms were thoroughly assessed during the USI A-46 and IPEEE walkdowns for unrestrained or inadequately restrained items that could move and potentially impact safety-related equipment. The greatest potential for stored items located inside the DG building would be in the hall area that is used as the primary entrance path to the DG and switchgear rooms. The A-46 and IPEEE walkdowns did not find any item, or combination ofitems, that had the potential to prevent access to the DG or switchgear rooms following a seismic event. However, in the unlikely event that access to the switchgear rooms was blocked through the primary entrance path, the switchgear rooms are directly accessible through the rear doors as described previously.
There is nothing in the DG building with the potential to cause humidity, temperature, or other adverse environmental condition to exist following a seismic event in excess of the normal conditions associated with DG operation other than the fire protection system. There are no potential steam line breaks in the DG building, and the room exhaust fans and louvers are included on the A-46 Safe Shutdown Equipment List. The fire protection system relays were evaluated for relay contact chatter as part of the A-46 evaluation. This evaluation showed that the fire protection system would not inadvertently actuate due to a seismic event.
Therefore, there are no credible barriers to successful operator performance to reset the DG relays.
HL-5377 E-4
l I
)
i ATTACHMENT Final Safety Analysis Report Change Form I and )
10 CFR 50.59 Evaluation !
l l
I l
l HL-5377
ATTACHMENT Final Safety Analysis Report Change Form and 10 CFR 50.59 Evaluation I
l l
I HL-5377
Form: FCF9401 A .
II ATCII NUCLEAR PLANT FSAR/FIIA Cil ANGE FORM ,
Organization O FHA Section No(s). or FCF No.: 14C-004 Assigned to:- SNC Unit No(s).: 1(2 IFSAR Section No(s).: See below .
INCORPORATED INTO SECTION(S) *IOCFR50.59 SAFETY EVALUATION DOCUMENTS ASSIGNED TO Tills FCF LOCATION 12.6 - 4f. / Attached FSAR Change 3 7 4 " M 2. gM/
Qy + M' TIIE EN ,OSED INFORM ION IS TRt!E, ACCIIRATE, CuntPl.ETE AND ADEQtJATELY St1PPORTED TO TIIE EXTENT OF SNC*? KNOWLEDGE.
(Nanee of Orpnization)
- 1. -
[O!/3 3 , have read arul followed the guidelines stated in ftwr. FCF9401B.
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- FSAR changes uhich are not editorial require a 10CFR50.59 Safety Evaluation. If a Safety Evaluation has been proiously performed, indicate the source (i.e., DCR, package, Procedure change, Technical Specification change, etc.). New Safety Evaluations, if requi cd, should be provided with the FCF response.
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HNP-1-FSAR-12 I N 12.6
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ANALYSIS OF SEISMIC CLASS 1 STRUCTURES 5 .
1
- 12.6.1 SCOPE '
i The loads, loading combinations, and allowable limits described r i here apply only to Seismic Class 1 structures. The criteria t i
are intended to supplement applicable industry design codes j where necessary to provide design safety margins for rare ;
i j
, events like postulated loss-of-coolant accident or earthquakes or tornadoes.
i The Seismic Class 1 concrete and steel structures are designed '
i considering 3 inter-related primary functions for tha design 1
loading combinations described in subsection 12.4. The first
{ consideration is to provide structural strength equal to or
{ greater than that required to sustain the combination of design
- loads and provide otection to other Seismic Class 1 i l
structures and nents. The second consideration is to maintain structura deformations within such limits that i Seismic Class 1 components and/or systems will not experience a ;
j loss of function. The third consideration is to limit
- excessive containment leakage by preventing excessive 4
(' deformation and cracking where containment integrity is required.
I i
12.6.2 STRUCTURAL ANALYSIS '
l In general, the structural analysis is performed utilizing the
, " working stress design" method as defined in American Concrete Institute (ACI) Standard Building Code Requirements for
( Reinforced Concrete (ACI 318-63), and in the American Institute i of Steel Construction ( AISC) Manual of Steel Construction (1963). Finite element stress analysis and other techniques
. are also used where applicable or necessary.
1
\. Load combinations and allowable limits on stresses are discussed in section 12.4. The maximum permissible calculated j
concrete compression is limited to 0.75 f'c , and the maximum permissible calculated main reinforcing steel tension is
! limited to 0.9 F . The maximum permissible calculated concrete j
( shearisasgive5inACI318-63, Chapter 17, for loadings j \ involving R and E'.
i i Bond and anchorage for reinforcing steel is treated as required 2 by ACI 318-63.
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12.6-1
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HNP-1-FSAR-A
) k'1 k A.3 DESIGN REOUIREMENTS I
l A.3.1 PIPING DESIGN Pressure and temperature conditions to which piping pressure 1 components are subjected are described in the appropriate I system design section of the final safety analysis report (FSAR). All piping systems within the scope of this appendix including pipe, flanges, valves, and fitting meet the requirements of American National Standards Institute (ANSI) s B31.1, or ANSI B31.7 as indicated in tables A.2-2 and A.2-3, l l including requirements for design, erection, supports, tests, l inspection, and special additional supplementary requirements l J
specified in this appendix.
A.3.1.1 Allowable Stresses The allowable stress values of the applicable piping code are used. For materials not covered by the piping codes, the stress values of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code are used.
A.3.1.2 Wall Thickness l Pipe wall thickness, fittings, and flange ratings are in accordance'with the applicable code, including adequate allowances for corrosion and erosion according to individual system requirements for a design life of 40 years.
A.3.1.3 Reactor Vessel Nozzle Lead All piping including instrument piping connecting to the reactor pressure vessel (RPV) nozzles is designed so that the ~
nozzle to pipe interface load does not result in stresses in
,) excess of the allowable material stresses. Thermal sleeves are used where nozzles are subjected to high thermal stresses.
Seismic Design n 56" A . 3 .1. 4 For ths purpose of seismic design, equipment and piping is categorized acco.iding to the following definitions:
Seismic Class 1 This class includes equipment and piping systems whose failure or malfunction could cause, or increase, the l
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A.3-1 REV 2 744
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Th 7he criteria for determining the seismic adequacy of mechanical and electrical equipment.for.
I
%,h Pkat He:h, UA ' a,w dcocubcd l m ibis 3cci en and c:her a = cf th: FS *S. for specMc Tim The criteria ar or analyses, Tha:remain rn::rivalid gusually specified and'may continue to be usedinasgeneral determined terms to be to include vei appropriate. However, as an alternative, ttie methodology for using the earthquake expenence data developed by the Seismic Qualificatio'n Utility Group (SQUG) and documented in the Generic Implementation Procedure (GIP))or use in resolving Unresolved Safety Issue (USI)
A-46jas required by NRC Generic Letter 87-02jnay be used to verify the seismic adequacy of currently installed equipment,as well as new and replacement mechanical and electrical
' equipment within the' scope eHhih T'e iacturien ef{his alternate method of verifying the seismic adequacy of equipment by :hi; ;m=d==: applies f8all seismic plasses of mechanical and electrical equipment. However, this attemate method of verifying the seismic adequacy of equipment used for modifications and replacement equipment assemblies, subassemblies, and devices that are part of the assemblies dea uvi suivinabwany supersede any eycdf= W car : -e-+e made te the MC (e g., per Reg. Guide 1.97 ar TMI Actica Itain II.".2).
T r hc, c-~h-e~r !c vedfj ::!:mb 2dequacy pc IEEE 2 " 1975 (reihe, ihau !""" 344- -
1971) =nn: be reph:cd by 'his gencre "SAR enmud.ncut te givvids fui ihc u;; cf thc_
SQUC CIP me:hede!cgy. %e phat; 3cnc;el:y commi;:cd :c IEEE : " 1975 and wcic 12ot nun,a,a; nth,.rinr,snh,,a ce m y n u:(Us!) a :S ca: gary orp:2,g , s g e;r,c cc m aments enr n,us!!6fng equipme" 6c!uding ec~~hme-s to m er lu m S,mu:tcentino,;vL, seNSrivd [va dic egt.jp mem for WRich thc Lumuunncni5 WC0 made Un'"'he Ormmitmente are I
_ C b.1..g;d b. awwvidauLc n Ads ibs ayysvyouic $5C;L^,i"'g b?? f ""Am"' ""^^*** l Si f "M dewit anA cq auHiacs hC5elv eg y is ae.eepf- ahle wkece no 1
spe ci4J NRc, 0 cnua;f ru,tf 40 us I rt66 Hy - ty 7f kas bren. /had.c. . -
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HNP-1-FSAR-A hYY p $fy/gp severity of the design basis accident (DBA), cause release of radioactivity in excess of 10 CFR 100 limits, or those essential for safe shutdown and immediate or long-term operation following a loss-of-coolant accident (LOCA).
Seismic Class 2 l l l This class includes equipment and piping system whose failure would not result in the release of significant radioactivity and would not prevent reactor shutdown. The failure of Seismic Class 2 equipment and piping systems may interrupt power generation. -
The equipment and piping considered as Seismic Class 1 are shown in table A.3-1. Seismic Class 1 equipment and piping systems are supported and restrained to meet the seismic design analysis criteria in compliance with applicable codes.
The dynamic analysis of Seismic Class 1 piping systems for seismic loads was performed using the spectrum response method, as applied to a lumped mass mathematical model of the piping systems. The maximum responses of each mode were calculated and combined by the square-root-of-the-sum-of-the-squares method to give the maximum response quantities resulting from all modes.
The response thus obtained was combined with the results produced by other loading conditions to compute the resultant stresses. All modes having frequencies less than 30 Hz are used. The percentage of critical damping used in the seismic analysis is defined in paragraph 12.3.3.2.1.2. The horizontal acceleration spectrum curves applied to the piping systems are developed as part of the seismic analysis for the building in which the piping is located.
A.3.1.5 Analysis of Piping A.3.1.5.1 Primary Stresses (Sp) (
Primary stresses are as follows:
A. Circumferential primary stress (SR) - Circumferential primary stresses are below the allowable stress (Sh) at the design pressure and temperature. ,
B. Longitudinal primary stresses (SL) - The following loads are considered as producing longitudinal primary b stresses: internal or external pressures; weight loads k.
A.3-2 REV 3 7/85
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j9L)U o f 4 Y ff SUPPLEMENT 3.7A d
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- g_ SEISMIC DESIGN el#
3/7/94 This section describes the seismic design requirements and methods used for Hatch Nuclear Plant-Unit 2 (ENP-2), and the seismic design and analysis of nonnuclear steam supply system equipment. Seismic design of nuclear steam supply system (NSSS) equipment is described in supplement 3.7B.
() 3.7A.1 SEISMIC INPUT The two types of seismic inputs used in the seismic analyses were the ground design spectra and the associated synthetic accelerogram.
3.7A.l.1 DESIGN RESPONSE SPECTRA Ground design spectra were established through extensive idvestigations on the geologicci conditions of the plant site and past seismological history of the neighborhood areas. The details of these investigations and the resulting
~' s recommendations are presented in section 2.5. The
.! recommendations were given in the form of maximum horizontal acceleration values of the ground, 0.08 g and 0.15 g for operating basis earthquake (OBE) and design basis earthquake (DBE), respectively. The modified Newmark design spectra associated with these acceleration levels were adopted and are shown in figures 3.7A-1 and 3.7A-2. They are characterized by a maximum amplification factor of 3.5 for 2 percent of critical damping and no amplification for frequencies beyond 30 Hz. l 3.7A.l.2 SYNTHETIC TIME HISTORIES l g 3.7A.l.2.1 Modified TAFT Time History l The synthetic acceleration time history shown in figure 3.7A-3 was developed for use as input to the time history analyses that resulted in the generation of the floor response spectra (FRS)
__) used to seismically qualify subsystems until April 4, 1985.
In developing this synthetic accelerogram, the first 20 s of the TAFT 1952 horizontal earthquake component was selected as the REV 2 7/84 3.7A-1 REV 3 7/85l
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Insert d b b pie e d e d. accer + M m- f s,
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Supplements 3.7A and 3.7B describe the seismic design requirements used to determine the seismic adequacy of mechanical and electrical equipmentJr.=Sd c.; ': ... !!-M, U.. ; 2- The 3gjg criteria :g:::f..d .'.. J mam ose,m... .d ed. . .,_.w..: :f ^^ N.LR for verifying the seismic
/
adequacy of equipment in a general manner or for specific equipment applications remain valid and may continue to be used as described. However, as an alternrtive, the methodology for using the earthquake experience data developed by the Seismic Qualification Utility Group (SQUG) and documented in the Generic Implementation Procedure (GIP)}for resolving Unresolved Safety Issue (USI) A-46 in response to NRC Generic Letter 87-02 may be used to verify the seismic adequacy ofcurrently installed equipmentgs well as new and replacement meenanical and electncal equipment wittun the scope of the GIP. This alternate method of verifying the seismic adequacy of equipment used for modifications and reolacement equipment assemblies, subassemblies, and devices that are a pan of the assembliesycer a^t eutc=t"h
"'mmcdm or ific ccrit-aa" m2de to the ".C (c ;;., ps. Reg. Guide 1.97 ction }
3:em 1 . her, commitments to verify seismic adequac 44-1975 (rather t tan IEEE 344-197 t be superseded by thi AR amendment to provide for the use of the SQUG GIP metho ants generally committed to IEEE 344-197!
, ahd were not included in the Up ed Sa ssue (USI) A-46 category of plants. Specific c ommitments for qualifying equipment, including commi o IEEE 344-1975, must c ontinue to be,satriffe'd e for the equipment for which the commitmenti'w until the chm nts are changed in accordance with the appropriate licensing basis ame M.. ,
a.Nw -lhs G. i ban akd debIneY n ad hM a odliers F.e solveA ,
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GF,ORGIA POWER COMPANY
. PLANT E.I. HATCil ,
FORM TITLE:
10 CFR 50.59 EVALUATION ,
O SHEET 1 OF 6 NUMBER: pc p. f jfc- ee y PROPOSED REVISION: O O TITLE: Verification of Seismic Adequacy of Equipment Assemblies, Subassemblies, and Desices That Are Part of the Assemblies DOCUMENT TYPE I- uni v a Unit I FSAR Amendment p 4, fg SYNOPSIS OF THE " ACTIVITY" TO WHICH THIS EVALUATION APPLIES:
6 As part of the resolutiun of USl A-46," Seismic Qualificat!on of Equipment in Operating Nuclear Power Plants," as required by NRC Generic Letter 87-02, the Seismic Qualification Utility Group (SQUG) collected data and developed criteria and procedures for verifying the susmic adequacy of nuclear plant i mechanical and electiical equipment. The methodology developed by SQUG is based primarily on the use of earthquake experience and generic test data, and is presented in the Generic Implementation !
Procedure (GlP). The GIP has been evaluated by the NRC and accepted for veriQing the seismic adequacy of plant equipment as documented in Supplement 2 to the SER (SSER No.2) on the GIP.
Justification for use of the GIP for new and replacement equipment assemblies, subassemblies, and devices that are part of the assemblies is provided by the following summary statements which are based on information contained in NRC SSER No. 2:
The USI A-46 (GIP) methodology is an acceptable evaluation method, for USE A-46 plants, to satisfy equipment seismic requirements of General Design Criterion (GDC) 2 and the purpose of NRC regulations relevant to seismic adequacy, including 10 CFR 100.
The USl A-46 methodology can be applied to new and replacement equipment within the scope of USI A-46.
USI A-46 plant licensees can change their seismic licensing basis to adopt the USl A-46 methodology for new and replacement items within the scope of USI A-46.
The USl A-46 methodology can be extended beyond he scope of the USl A-46 SSEL, and be applied to all plant electrical and mechanical equipment within the scope of the GIP.
6 The above siniements clearly indicate that the licensing basis contained in the FSAR can be changed to allow the use of the GIP methodology. However, additional seismic requirements imposed by the NRC or commi'.ted to by licensees in response to specific issues (e.g., Regulatory Guide 1.97 or TMI Action item II.F.2) are not superseded automatically by an amendment to the FSAR to adopt the GIP, the provisions of Gervric Letter 87-02, or SSER No. 2. Specific licensing commitments outside the FSAR must be formally sevised following applicable procedures and regulations. Commitments to IEEE 344-1975 are within this category of specific commitments and must be adhered to until appropriate actions are take to revise the commitments. That is because USI A-46 plants generally committed to IEEE 3441971 or less specific criteria for seismically qualifying %nt equipment. Newer plants that committed to IEEE 344-1975 were considered to have had adequate seismic review and were not required to perform the USl A-46 evaluation.
h i
ONCE & SCREENING QUESTION IS ANSWERED "YES", THE REMAINDER OF TIIE SCREENING OUESTIONS ARE NOT REQUIRED TO BE ANSWERED.
MGR-0020 REV. 2 N/A 10AC-MGR-010-0S
- GEORGIA POWER COMPANY f
. PLANT E.I. IIATCII FORM TITLE:
10 CFR 50.59 EVALUATION O SIIEET 2 OF 6 10 CFR 50.59 SCREENING (i.e., BLOCKS O AND O):
} (El YES ;O NO ;
- Is the " ACTIVITY" itself a change to one of the following, OR .Is a change to one of the following required as a result of the " ACTIVITY"
- a. the Technical Specifications and / or the Environmental Technical Specifications incorporated in the l Operating License, 0._R i
- b. Other licensing document (s) as defined in 00AC-REG-003-0S? !
i' BASIS FOR ANSWER:
( g The activity represents a change to the FSAR. The FSAR is described in 00AC-REG-003-OS. Therefore, i the activity is a change to a document defined in 00AC-REG-003-OS.
I i
1 I
k i
] gy E the answer is "YES", complete the CONTROL OF CilANGES TO LICENSING DOCUMENTS form, i W4 AND make it a part of the 10 CFR 50.59 EVALUATION packaec.
I 10 CFR 50.59 SCREENING (CONTINUED): i 4 l E APPLICABLE / DESIRED, GO DIRECTLY TO A QUESTION TIIAT IIAS A "YES" ANSWER 1
- Does the " ACTIVITY" to which this evaluation applies represent
BASIS FOR ANSWER:
1
- 2. O YES O NO A change to procedures described in the FSAR?
BASIS FOR ANSWER:
MGR-0020 REV. 2 N/A 10AC-MGR-010-OS
- i i GEORGIA POWER COMPANY
. PLANT E.1. IIATCil FORM TITLE:
, 10 CFR 50.59 EVALUATION O SHEET 3 OF 6 l O 4
l j zI l 3. O YES O NO A test or experiment not described in the FSAR?
1 BASIS FOR ANSWF.R:
l i
l IF the answers to ALL the questions in Blocks O and O are "NO " complete Blocks O through O.
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[F the answer to ANY question in Blocks O and O is "vES," complete Biocks O through O. l PREPARED: b- DATE: /c //f/ 9 5 6 REVIEWED: c, #_ DATE: // f) /jis APPROVED: 4f
- DATE: // //% fS l
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1 MGR-0020 REV. 2 N/A 10AC-MGR-010-OS
. GEORGIA POWER COMPANY
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PLANT E.1. IIATCil i
FORM TITLE:
10 CFR 50.59 EVALUATION O SilEET 4 OF 6 SAFETY EVALUATION
- 1. O YES (Z) NO Does the proposed " ACTIVITY" increase the probability of occurrence of an ;
accident previously evaluated in the FSAR.
BASIS FOR ANSWER:
This activity represents a change to the FSAR to provide for the use of earthquake experience data as an alternate method for verifying the seismic adequacy of new and replacement equipment. The purpose of this evaluation is to determine whether of not the proposed change the FSAR constitutes an unresolved safety question. The NRC has reviewed the methodology for use of the earthquake experience data in i accordance with the guidance of the Seismic Qualification Utility Group (SQUG) Generic implementation !
Procedure (GIP) and determined that it is acceptable for verifying the seismic adequacy of mechanical l and electrical equipment in USI A-46 nuclear power plants. That NRC acceptance of the SQUG i methodology is documented in the NRC Safety Evaluation Report, Supplement 2 (SSER No. 2), dated l 4
May 22,1992, which also allows the use of the methodology for verifying the seismic adequacy of new and replacement equipment. Hatch Nuclear Plant (HNP) was required by Generic Letter 87-02 to resolve USl A-46 in accordance with the guidance of the GIP. Therefore, Hatch is a USl A-46 plant. Since the use of the earthquake experience data has been reviewed and approved by the NRC, an unresolved safety question associated with its use at HNP does not exist. Therefore, the proposed activity does not increase the probability of occurrence of an accident previously evaluated in the FSAR.
- 2. O YES El NO Does the proposed " ACTIVITY" increase the (radiological) consequences of an accident previously evaluated in the FSAR?
BASIS FOR ANSWER:
O The proposed activity provides for an altemate method for assuring that plant equipment is seismically adequate to withstand the effects of a design basis earthquake. It does not modify any plant equipment or change the way the equipment is operated, and it does not alter any radiological release control or mitigation functiont.. 'Therefore, the activity does not increase the radiological consequences of an accident previously evaluated in the FSAR. l
- 3. O YES S NO Does the proposed " ACTIVITY" increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR?
BASIS FOR ANSWER:
The proposed change does not modify any plant equipment or structures. It does not change the way the equipment is operated. It provides for an alternate method for verifying the seismic adequacy of plant equipment. Since the alternate method is equivalent to or better than the current method,it does not increase the probat'!!;iy of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR.
- 4. O YES S NO Does the proposed " ACTIVITY" increase the (radiological) consequence; of a malfunction of equipment important to safety previously evaluated in the FSAR?
BASIS FOR ANSWER:
MGR-0020 REV. 2 N/A 10AC-MGR-010-OS
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G.EORGIA POWER COMPANY
, . PLANT' E.I. IIATCll VORM TITLE:
10 CFR 50.59 EVALUATION O SHEET 5 OF 6 SAFETY EVALUATION l l The proposed FSAR change does not modify any plant equipment or structure or change the way the equipment is operated. It does not alter any radiation release control or mitigating functions. Therefore, the proposed actisity does not increase the radiological consequences of a malfunction of equipment important to safety presiously evaluated in the FSAR.
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- 5. O YES El NO Does the proposed " ACTIVITY" create the possibility of an accident of a different type than any previously evaluated in the FSAR?
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- BASIS FOR ANSWER: j l The proposed activity does not modify any plant equipment or change plant procedures used to operate the
, equipment. It only provides an alternate method for verifying the seismic adequacy of the equipment to withstand a i design basis carthquake and continue to perform its intended function. That method of verifying the seismic adequacy of equipment was developed by SQUG in the form of the GIP. It was used to evaluate the adequacy of
, equipment installed in accordance with current licensing basis for seismically qualifying the equipment. Since the i SQUG methodology was required in addition to the current licensing basis method and as an evaluation of the ;
] adequacy of the current licensing basis method, it follows logically that it is at least equivalent to the current method 1 for assuring the seismic adequacy of the equipment. The proposed FS AR change does not modify plant equipment or change the way the equipment is operated, and it provides for an equivalent or superior method for verifying the seismic adequacy of the equipment. Therefore, the proposed activity does not create the possibility of an accident of a different type than any previously evaluated in the FS AR.
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- 6. O YES E NO Does the proposed " ACTIVITY" create the possibility of a malfunction of
< equipment important to safety of a different type than any previously evaluated in
- the FSAR?
i l BASIS FOR ANSWER:
O The proposed change to the FSAR does not modify any plant equipment or structures. It does not change the way plant equipment is operated. The proposed change only provides an alternate method for verifying the seismic ;
j adequacy of the equipment. That alternate method is equivalent or superior to the current licensing basis method.
1 Therefore, the proposed activity will not create the possibility of a malfunction of equipment important to safety of a
- difTerent type than any presiously evaluated in the FSAR.
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- 7. O YES ENO Does the proposed " ACTIVITY" reduce the margin of safety as defined in the basis for any Technical Specification?
BASIS FOR ANSWER:
The proposed activity would revise the FSAR to provide for the use of the SQUG GIP methodology as an alternate method for verifying the scismic adequacy of plant equipment. The GIP methodology has been evaluated and approved by the NRC for verifying the adequacy of mechanical and electrical equipment in USI A-46 plants. That NRC approval has been documented in SSER No. 2 for currently installed equipment as well as new and replacement equipment. IIence, the inclusion of the GIP methodology in the licensing basis for the seismic venfication of equipment does not reduce the margin of safety since regulatory requirements continue to be satisfied and the radiological consequences of a postulated design basis accident remain within regulatory limits. An unreviewed safety question does not exist based on the above discussions and on the fact that the activity has been reviewed by NRC. Therefore, the proposed activity will not reduce the margin of safety as defined in the basis for any Technical Specification.
MGR-0020 REV. 2 N/A 10AC-MGR-010-OS
1 GEORGIA POWER COMPANY l ,
PLANT E.1. IIATCil FORM TITLE:
10 CFR 50.59 EVALUATION O SHEET 6 OF 6 SAFETY EVALUATION E a change to the Technical Specifications or the Environmental Technical Specifications is required,9R E ANY
, of the questions in Block O is answered "YES." an unreviewed safety question IS Indicated. In that case, approval from the NRC is required BEFORE the " ACTIVITY" can be implemented. Refer to subsection 8.5.1.2 for guidance on exceptions to this, i
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4 MGR-0020 REV. 2 N/A 10AC-MGR-010-0S l