ML20236R867
| ML20236R867 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 07/09/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20236R857 | List: |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR GL-87-02, GL-87-2, NUDOCS 9807220328 | |
| Download: ML20236R867 (11) | |
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UNITED STATES i
j NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 20606 4 001 l
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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO USI A-46 PROGRAM IMPLEMENTATION SOUTHERN NUCLEAR OPERATING COMPANY. INC.. ET AL.
JOSEPH M. FARLEY NUCLEAR PLANT. UNIT NO.1 DOCKET NO. 50-348
1.0 BACKGROUND
i On February 19,1987, the NRC issued Generic Letter (GL) 87-02, " Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety issue (USI) A-46." The GL encouraged licensees to participate in a generic program to resolve the seismic verification issues associated with USI A-46. As a result, the Seismic Qualification j
Utility Group (SQUG) developed the " Generic implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment," Revision 2 (GIP-2) (Reference 1).
On May ~,1992, the NRC issued Supplement 1 to GL 87-02 including the staffs Supplemental Safety Evaluation Report No. 2 (SSER-2) (Reference 2), pursuant to the i
provisions of 10 CFR 50.54(f), which required that all addressees provide either (1) a l
commitment to use both the SQUG commitments and the implementation guidance described in GIP-2 as supplemented by the staffs SSER-2, or (2) an attemative method for responding to GL 87-02. The supplement also required that those addressees committing to implement GIP-2 provide an implementation schedule as well as detailed information including the procedures and criteria used to generate the in-structure response spectra (IRS) to be used for USl A-46.
l By letter dated September 10,1992 (Reference 3) Southern Nuclear Operating Company, Inc.
(SNC) provided its response to Supplement 1 to GL 87-02 for Joseph M. Farley Nuclear Plant, Unit 1. In that letter, SNC committed to follow the SQUG commitments set forth in GlP-2, including the clarifications and exceptions identified in the staffs SSER-2. The staffs evaluation of SNC's response was issued in a letter dated November 20,1992 (Reference 4).
Note that the USl A 46 program is not applicable to Unit 2 because Unit 2 is not a USl A-46 plant.
SNC conducted the USl A-46 program and submitted a summary report on May 18,1995 (Reference 5). The staff reviewed the summary report and requested additionalinformation (RAI) on September 9,1995 (Reference 6), August 29,1996 (Reference 7), and May 15,1997 (Reference 8). SNC responded to the staffs RAls on October 11,1995 (Reference 9),
Enclosure 9807220328 980709 i
PDR ADOCK 05000348 i
P POR t
6 October 28,1996 (Reference 10), and August 11,1997 (Reference 11). The staff has completed its review of all of SNC's responses mentioned above.
In its submittal of May 18,1995 (Reference 5), SNC stated its intention to revise the licensing basis for Unit 1 and Unit 2 to allow application of GlP-2 methodology for qualifying mechanical and electncal equipment in both units. In early 1996, SNC made a Final Safety Analysis Report (FSAR) change for Unit 1 in accordance with the provisions of 10 CFR 50.59 on a program
- level. In various discussions with SQUG relating to generic USl A-46 issues, the NRC staff
- expressed its concerns regarding the revision of the licensing basis for a USI A46 plant in accordance with the provisions of 10 CFR 50.59 on a program level without addressing all aspects where incorporation of GlP-2 may introduce the use of criteria not in compliance with the licensing basis. The use of GIP-2 is not appropriate for non-USl A-46 plants as stated in SSER-2 (Reference 2). In response to the stars RAl, SNC stated, in Reference 10, that it would not apply the GIP-2 methodology at this time to Unit 2 for new and replacement equipment. Furthermore, SNC stated, in its response to an RAI item in Reference 11, that SNC initiated another FSAR change to withdraw the previous change that would have allowed the l
use of earthquake experience data for verifying the seismic adequacy of new and replacement equipment for both FNP Units 1 and 2. SNC indicated that any application of the SQUG GlP-2 methodology will be implemented on a case-by-case basis with NRC staff approval.
This report provides the staff evaluation of SNC's USI A 46 implementation program at Unit 1 based on the stars review of SNC's summary report (Reference 5), and documentation provided by SNC in response to the stars RAls.
l 2.0 DISCUSSION AND EVALUATION The summary report (Reference 5) provides the SNC's implementation results of the USl A-46 program at Unit 1. The report contains safe shutdown equipment identification, seismic screening verification and walkdown of mechanical and electrical equipment, the relay evaluation, seicmic adequacy of tanks and heat exchangers, seismic adequacy of cable and conduit raceways, and outlier identification and resolutions.
2.1 Seismic Demand Determination (Ground Spectra and in-structure Response Spectra (IRS))
Unit 1 was designed using a modified Newmark spectrum with a horizontal peak ground acceleration of 0.10g and a vertical peak ground acceleration of 0.067g for the safe shutdown earthquake (SSE). IRS at key elevations of Unit 1 structures including the auxiliary and containment buildings were provided in SNC's letter dated October 28,1996 (Reference 10). In its letter dated September 10,1992 (Reference 3), SNC provided a response to NRC's Supplement No.1 to GL 87-02 (Reference 2) and committed to use, for the implementation of its USl A-46 program, the plant's originallicensing basis IRS for structures, systems, and components located within the auxiliary building, containment internal structure and equipment supported by the containment shell. The staff finds this acceptable.
SNC stated that the diesel generator building (DGB) and service water intake structures (SWIS) were supported on cast-in-place caissons which transfer the loads to the competent underlying l
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rock material, and that the simplified modeling of the caissons in the generation of the original IRS as uescribed in the plant FSAR was outdated. Therefore, SNC generated new IRS for these two structures. The new IRS was developed in accordance with the 1989 revision of the NRC Standard Review Plan. SNC used the CLASSl/ SHAKE computer programs in the l
generation of the new IRS. The resulting new response spectra in the free field at the DGB and SWIS foundation elevations were judged by SNC as satisfying the 10 CFR Part 100 Appendix A requirement that the maximum ground motion is no less than 0.10g. The staff found these new IRS acceptable for use in addressing USl A-46 at Unit 1.
l 2.2 Seismic Evaluation Personnel SNC's summary report (Reference 5) states that the Seismic Review Team (SRT) members were the personnel t r>.,ponsible for applying experience and judgement in the implementation of l
the combined Electric Power Research Institute (EPRI) seismic margin assessment (SMA) i methodology and SQUG GIP 4. Work experience and training for SRT members for Unit 1 L
included many years of formai and practical experience in the field of structural and seismic design and analysis. The combined qualifications of the SRT members met the provisions of l
the SQUG GlP-2. Each SRT member attended the SQUG Walkdown Screening and Seismic Evaluation Training Course.
The summary report also states that the " Third Party" review was conducted by Dr. John Reed of Jack R. Benjamin and Associates. Dr. Reed is a key author of the final version of the EPRI i
l SMA methodology Ed taught the Add-On Seismic Individual Plant Examination Training l
Course with Dr. Robert P. Kennedy. Dr. Reed's audit of the USl A-46 program implementation l.
at Unit 1 consisted of a 1-day plant walkdown, a follow-up meeting, and a review of the Unit 1 l
USl A-46 summary report. Dr. Reed concluded that SNC's USl A-46 program implementation has satisfied the provisions of SQUG GlP-2 (Appendix L to Reference 5).
i The staff finds that SNC's seismic evaluation personnel qualifications meet the provisions of GlP-2 snd the staff's SSER-2, and are, therefore, acceptable.
2.3 Safe-Shutdown Path l'
GL 87-02 specifies that the licensee should be able to bring the plant to, and maintain a hot shutdown condition during the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an SSE. To meet this provision, in its submittal of May 18,1995 (Reference 5), SNC addressed the following plant safety functions:
reactivity control; inventory control; and decoy heat removal. A primary and attemate path were identified for each of these safety functions to ensure that the plant is capable of being brought to, and maintained in a hot shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an SSE.
Short-tsrm reactivity control is achieved through insertion of control rod assemblies (reactor trip). Long-term reactivity control and reactor ce jant inventory control are achieved through injection of borated water from the refueling wrner storage tank (RWST) to the reactor coolant system (RCS). Injection is achieved via the safety-related charging pumps and associated flow paths. Reactor decay heat removal is accomplished by two sequential methods. The first method is secondary heat removal. These operations consist of feeding water from the condensate storage tank (CST) or the service water (SW) system to the steam generators l
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~ (SGs) via the safety-related motor driven auxiliary feedwater pumps and allowing steaming i
(boiling off) of the SGs through the main steam atmospheric dump valves. The second method l
is by using the residual heat removal (RHR) systcm in the decay heat removal mode of opershon. This method is used after the secondary heat removal mode brings the RCS temperature and pressure to approximately 350 *F and 425 psig, respectively. In this method, l
decay heat is removed from the reactor coolant to the component cooling water (CCW) system l
by way of the RHR heat exchangers The heat in the CCW system is then rejected to the SW system by the CCW heat exchangers.
SNC has identified several diverse water sources to ensure that the plant safety functions are.
maintained. The seismically qualified water sources are the CST and the RWST, which have capacities of 500,000 gallons each, respectively. The SW system provides a backup source of water to the auxiliary feedwater system should the contents of the CST be depleted. The SW j
system compnses a 100-acre storage pond with a seismically qualified dam, pump house, and distribution piping. All SW pumps and auxiliary equipment are provided with Class 1E power supplies.
Other nonseismically qualified sources of water, which are not included in the SSE list (SSEL) are available for makeup to the primary and secondary systems. These are:
l a.
Seismic Primary Makeup Water Sources:
Reactor Makeup Water Storage Tank 200,000 gallons Two Boric Acid Storage Tanks 21,000 gallons j
b.
Nonseismic Secondary Makeup Water Source:
Demineralized Water Storage Tank 200,000 gallons in the unlikely event that all of these cooling sources are unavailable, the operators would implement the primary feed and bleed mode of cooling that uses the RWST and the high head safety injections pumps. The components and equipment within this cooling path are seismically qualified.
A composite list of the equipment necessary for the above-mentioned paths is included in Appendix A of Reference 5. The components contained in this list were walked down, and their j
seismic adequacy verified by SNC. In addition, SNC performed reviews and verified the -
l adequacy of the SSEL and plant operating procedures, and the compatibility of the SSEL with
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l these procedures.
The primary and altomate success paths are completely independent of each other. This is accomplished by using separate power supplies, different systems, opposite systems trains, and redundant components within systems whenever possible, in each of the paths, to perform the safety functions. The numerous water sources, flow paths, and time available for achieving i
the operations alignments, provide reasonable assurance that adequate decay heat removal t
capability is available. Symptom-based abnormal and emergency op3 rating procedures for lining up cooling sources are available at the site. The staff concludes that SNC's approach to achieve and maintain hot shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after a postulated seismic event is acceptable.
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l l i 2.4 Seismic Screening Verification and Walkdown of Mechanical and Electrical Equipment The seismic screening and walkdown included verification of 498 equipment items in the 20 classes of equipment included in GIP-2 and the "other" equipment class not covered in j
GlP-2 (i.e., neutron flux monitoring assemblies, tanks, and neat exchdrgers).
2.4.1 Equipment Seismic Capacity Compared to Seismic Demand i
The summary report (Reference 5) states that a complete list of these components is in the j
seismic review SSEL, which is included in Appendix H. Each component was evaluated for
. seismic capacity versus demand, conformance to caveats, anchorage adequacy, and seismic interaction effects. The results of the seismic capability walkdown for each SSEL component are documented on the Screening Verification Data Sheets (SVDS), which is included as Appendix G.
l SNC stated in the summary report (Reference 5) that all SSEL equipment at Unit 1 is located at an elevation less than about 40 feet above grade. Method A of Table 4-1 of the GlP-2 l
(Reference 1) was used to address the seismic adequacy of equipment located at an elevation j
less than approximately 40 feet above the grade of the auxiliary building and the containment i
l intamal structure with a natural frequency above approximately 8 Hz. In its response (Reference 10) to the staff's RAI, SNC indicated that some licensing basis IRS applicable to the t
containment intemal structure and the auxiliary building slightly exceed 1.5 times the GlP-2 Bounding Spectrum. This was because of the large conservatism inherent in the development of the original design IRS. Therefore, the staff found that SNC's application of Method A of Table 4-1 was consistent with the applicable GIP-2 (Reference 1) guidance and pertinent NRC position for the resolution of the USl A-46 issue, and, therefore, was an adequate way of comparing seismic capacity to seismic demand for Unit 1 plant equipment at an elevation less than about 40 feet above grade and with frequencies greater than 8 Hz.
Based on info mation provided in Appendix G to the Unit 1 summary report (Reference 5) and SNC's responses to the staffs RAI (References 10 and 11), most of the seismic capacity l
versus seismic demand comparisons for equipment in the SSEL was implemented by l
comparing the GlP-2 Bounding Spectrum against the ground response spectrum with some exceptions. Other approaches, including use of component-specific seismic qualification i
documentation, and comparison of 1.5 times the Bounding Spectra versus the seismic demand (IRS), were also adopted in assessing the adequacy of equipment seismic capacity. These approaches adopted by SNC in assuring that equipment seismic capacities are greater than l
their corresponding seismic demands are judged to be consistent with the guidelines provided in GIP-2 and the staffs positions described in SSER-2, respectively, and are, therefore,
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acceptable.
I 2.4.2 Assessment of Equipment" Caveats" in order to apply experience-based approach and use the equipment seismic capacity defined in GIP-2, the plant-specific equipment must meet some restrictions or caveats described in GIP-2. GIP-2 also allows engineers to verify whether the plant equipment conditions satisfy the caveats specified for a particular equipment class by judging whether these conditions meet the i
- " intent of the caveats" and not necessarily the words of the caveats. The summary report (Reference 5) states that to_ simplify the Unit 1 walkdown, components that did not meet the specific wording of the caveat were usually shown as outliers that were then resolved by analysis or other appropriate means to document that the intent of the caveat was met. If a component did not meet the wording of a caveat, and was not found as an outlier, the justification for meeting the intent of a caveat was noted on the Screening and Evaluation Work Sheets. The results were indicated in Column 13 of Appendix G to the summary report. The staff finds SNC's approach for assessing the equipment " caveats" to be reasonable and acceptable.
2.4.3 Equipment Anchorage i
SNC verified equipment anchorage during the walkdown and documented the acceptability of equipment on the SVDS as shown in Appendix G of Reference 5.
SNC used the SSE ground response spectra multiplied by 1.5 times 1.25 (1.875) to define the i
seismic demand for anchorage evaluation of equipment with a fundamental frequency greater
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than or equal to approximate y 8 Hz. This is consistent with the guidance given in Table 4-3 of GIP-2. For equipment in the auxiliary and containment buildings with a fundamental frequency i
of less than approximately 8 Hz, the original licensing basis IRS was used to determine the seismic demand for anchorage evaluation. The new IRS generated for the diesel generator building and service water intake structure were used for equipment located 'in these buildings with an additional factor of safety of 1.25 applied over those given in Table 4-3 of GIP-2. The methods used to determine seismic capacity of equipment anchorage followed the procedures provided in Appendix C to GlP-2. These methods were previously reviewed and accepted by the staff for assessment of anchorage capacities as a part of the USl A-46 evaluation.
l Among some 65 equipment outliers identified, approximately 30 percent of the outliers were related to anchorage deficiencies including anchor bolt corrosion, inadequate anchorage, potential bolt bending concoms, inadequate bolt torque resistance, inadequate anchorage and load path, and missing anchor bolts. These anchorage-related outliers were identified in Appendix K to the summary report and were resolved by bolt replacement, anchorage modification, or engineering analysis.
The above-described approaches used by SNC in ensuring adequate equipment anchorage capacity to withstand seismic effects are consistent with the staff position as discussed in SSER-2 and the applicable GIP-2 guidelines, and therefore, are judged as an adequate means for resolving the USl A-46 anchorage issue.
2.4.4 Seismic Spatial Interaction Evaluation Approximately 4g8 components were evaluated during the Unit 1 USl A-46 plant walkdown.
Among the key aspects evaluated during the walkdown was seismic capacity versus demand, conformance to caveats, anchorage adequacy, and seismic interaction effects. The results of the walkdown were tabulated in the SVDS. Some 50 components were identiiied to have potential adverse interaction effects. The method used in assessing whether equipment is free of interaction effects, or the interaction effects are acceptable and do not compromise the safe shutdown function of the equipment was evaluated and judged to be adequate for resolving the
. spatial interaction issue. The outliers identified were resolved by installation of additional restraints to the vulnerable equipment items. Therefore, the staff finds SNC's seismic spatial interaction evaluation acceptable for the resolution of USl A-46.
2.5 Tanks and Heat Exchangers A total of nine heat exchangers and two flat-bottom, cylindrical vertical tanks were evaluated by SNC. The summary repost (Reference 5) stated that these tanks and heat exchangers were evaluated for tank wall adequacy, anchorage adequacy, adequacy of connections between anchor bolts and tank shells, and flexibility of attached piping in accordance with GlP-2 criteria.
The results of the evaluation are summarized in Appendix G of the summary report.
The vertical tank evaluations include the CST and the RWST. The evaluation of the CST was based on the methodology provided in Section 7 of GlP-2. Based on the performance of seismic walkdown and associated engineering calculations, SNC determined that the tank shell capacity exceeded the demand, and that the other aspects of the tank including the anchorage, the attached piping, and the foundation were adequate for assuring tank integrity during and after an SSE. The RWST is a 500,000-gallon flat-bottom, vertical water storage tank located at plant grade (elevation 154.5 feet). New ground response spectra were developed by SNC for the seismic demand of tanks mounted on the ground suiface in the plant yard. These seismic demand spectra were established by performing three analyses using three sets of high strain soil properties. The IRS from these three best estimate analyses was then enveloped and peak broadened by *10% to yield the demand spectra for the tank. The seismic adequacy of the RWST tank was evaluated using the methodology provided in Section 7 of GIP-2. By use of this more-refined definition of the seismic demand, the evaluation resulted in a factor of safety of 1.04 against overtuming. Considering the conservatism incorporated in defining the seismic demand and in the tank design, the resulting safety factor of 1.04 for an existing tank is judged to be adequate. Among heat exchangers reviewed are Spent Fuel Pool, Residual Heat Removal, Seal Water, and Letdown heat exchangers. The review also covered CCW Surge tank, Fuel Oil Day tank, Boron injection tank, and Fuel Storage tank. SNC performed seismic capacity walkdowns of these tanks and heat exchangers using the guidelines provided in Section 7 of GIP-2, and verified their seismic adequacy.
The staff finds SNC's evaluation of tanks and heat exchangers acceptable for the resolution of USl A-46 at Unit 1.
2.6 Cable and Conduit Raceways SNC implemented the review of cable and conduit raceway supports in two parts: (1) a plant walkdown that evaluated the raceways and supports following the GIP-2 inclusion rules and walkdown guidelines, and (2) an analytical check of selected worst-case supports using
- Section 8 of GIP-2 for limited analytical review guidelines. The seismic demand for cable tray and conduit raceways were determined based on the methodology discussed in Section 2.1 of this SE. The raceway solustion program was carried out by SNC following the guidelines of GlP-2 and the
- Review Procedure to Assess Seismic Ruggedness of Cantilever Bracket Cable Tray Supports"(Reference 8).
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. Some 15 typical crsbie tray and raceway systems located in Units 1 and 2 (SNC's report was written with Unit 2 included) were selected for the limited analytical review, of which nine cases are applicable to Unit 1. Of the 15 systems reviewed, five systems located in Unit 1 were identdied as outliers. Two of the outliers were resolved with modification of base plate and support elements whereas the remainder of the outliers was resolved by more refined engineering analyses. Appendix J of the summary report (Reference 5) lists the results of the
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cable tray and conduit raceway supports evaluation. The staff finds that SNC's approach in verifying seismic adequacy of cable and conduit. raceway systems is adequate and acceptable.
2.7 Essential Relays
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SNC identified, in the summary report (Reference 5), three electrical engineers who were responsible for the relay chatter evaluation and identification of electrical SSEL components.
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The three individuals possess considerable combined experience in the design and analysis of electrical systems for nuclear power plants. SNC used the criteria and assumptions given in Section 3.3.1 of the EPRI-7148-SL report (Reference 13) for the relay chatter evaluations, which is consistent with GIP-2 procedures. A relay walkdown was performed for essential relays to ensure seismic adequacy of the cabinets or enclosures, which support the essential relays, and to spot check mountings of essential relays. SNC used Generic Equipment Ruggedness Spectra and component-specific seismic qualification for determining the seismic capacity of relays for comparison with the seismic demand in accordance with GlP-2 criteria and procedures.
SNC identified six relays of two different models as relay outliers. The seismic capacity data available for these relays is inadequate for the installations. Chatter of these devices could potentially result in unacceptable circuit actions due to a seal-in condition. The summary report (Reference 5) stated that corrections were required for these relay outliers. By letter dated December 20,1996, SNC stated that those necessary actions to resolve the outliers have been completed.
in its response (Reference 10) to the staffs RAI concoming " bad actor" relays, SNC stated that, in addition to the relays identified in Section 3 of Reference 5, GE PVD relays were identified in the safe shutdown path. Shake table testing meeting the requirements of IEEE 344-1975, demonstrated that these relays are acceptable for use at Unit 1. In Reference 9, SNC also indicated that no relay acceptance was relying on using operator action. The staff finds SNC's approach in verifying seismic adequacy of essential relays at Unit i reasonable and is,
. therefore, acceptable for USl A-46 resolution.
l 2.8 Human Factors Aspect r
l The staff's review focused on verifying that SNC had used one or more of GIP-2 methods for conducting the operations department review of the SSEL, and had considered aspects of human performance in determining what operator actit.,ns could be used to achieve and maintain safe shutdown (e.g., resotting relays, manual operation of plant equipment). The staff requested additional information regarding operator actions specified in SNC's summary report (Reference 5) in a letter dated January 2,1997. By letter dated August 11,1997 (Reference 1),
SNC provided the staff with a response to the RAl.
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.g.
SNC provided information that outlined the use of the " desk-top" evaluation method by a j
licensed senior reactor operator to verify that existing normal, abnormal, and emergency operating procedures were adequate to mitigate the postulated transient and that operators could place and maintain the plant in a safe shutdown condition. The staff verified that SNC
' had considered its operator training programs and verified that its training was sufficient to l
l ensure that those actions specified in the procedures could be accomplished by the operating crews. In addition, the staff requested verification that SNC had adequately evaluated potential challenges to operators, such as lost or diminished lighting, harsh environmental conditions,
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potential for damaged equipment interfering with the operator's tasks, and the potential for placing an operator in unfamiliar or inhospitable surroundings. SNC provided information regarding its seismic interaction evaluations and "detNop" evaluations to substantiate that i
operator actions could be accomplished in a time franta required to mitigate the transient.
Specifically, SNC provided assurance that ample time existed for operators to take the required l
actions to safely shut down the plant and described all SSEL equipment as located in Seismic Category 1 structures that have been demonstrated to remain intact with no structural damage I
that could hinder operator action.
SNC has provided the staff with sufficient information to demonstrate conformance with the review methodology outlined in GIP-2. Therefore, SNC's program related to the human factors i
aspects is acceptable.
I 2.9 Outlieridentification and Resolutions l
SNC identified six relays as outliers and the resolution of these relay outliers have been completed as described in 'Section 2.7 of this SE. Identified outliers for a cable and conduit raceway consist of two types: (1) outliers identified as a result of limited analytical review listed in Appendices I and J of the summary report (Reference 5), and are described in Section 2.6 of this SE, and (2) outliers identified as a result of seismic walkdown are described in Appendix J of the summary report, including the methods of outlier resolution.
Appendix K of the summary report contains a summary of all SSEL components identified as outliers during SNC's USl A-46 evaluation. A brief outlier resolution for each outlier item was provided in the Appendix. The staff's evaluation of outliers related to relays, equipment I
anchorages, spatial interaction, and cable and conduit raceway supports are discussed in respective sections in this SE. No outliers were identified by SNC for tanks and heat exchangers.
By letter dated December 20,1996, SNC stated that necessary actions to resolve all the outliers have been completed. The staff finds SNC's approach in identifying and resolving outliers reasonable and acceptable.
l 3.0
SUMMARY
OF MAJOR STAFF FINDINGS The staffs review of SNC's USI A 46 implementation program, as provided for each area discussed above, did not find any significant or programmatic deviation from GIP-2 regarding the walkdown and the seismic adequacy evaluation at Unit 1.
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4.0 CONCLUSION
SNC's USl A-46 program at Unit 1 was established in response to Supplement 1 to GL 87-02 through a 10 CFR 50.54(f) letter. SNC conducted the USl A 46 implementation in accordance with GIP-2 and the stars SSER-2. SNC's submittal on the USl A-46 Implementation indicated that a total of 498 components were evaluated during the Unit 1 seismic walkdown. SNC
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identified approximately 65 components as outliers. By letter dated December 20,1996, SNC j
l stated that necessary actions to resolve the outliers have been completed. SNC's implementation report did not identify any instance where the operability of a particuler system or component was questionable.
The staff concludes that SNC's USl A-46 implementation program has, in general, met the purpose and intent of the criteria in GlP-2 and the stars SSER-2 on GlP-2 for the resolution of USl A-46.. The staff has determined that SNC's actions will result in safety enhancements, in certain aspects, that are beyond the original licensing basis. As a result, SNC's actions provide sufficient basis to close the USl A-46 Review at the facility. The staff also concludes that its findings regarding SNC's implementation of USl A-46 do not warrant any further regulatory action under the provisions of 10 CFR 50.54(f). Licensee activities related to the USl A 46 implementation may be subject to NRC inspection.
Regarding future use of GIP-2 in licensed activities, SNC may revise its licensing basis in accordance with the guidance in Section 1.2.3 of the stars SSER-2 on SQUG/GlP-2 j
(Reference 2) and the stars letter to SQUG's Chairman, Mr. Neil Smith, on June 19,1998.
l Where plants have specific commitments in the licensing basis with respect to seismic l
qualification, these commitments should be carefully considered. The overall cumulative effect 1
l of the incorporation of the GlP-2 methodology, considered as a whole, should be assessed in l
making a determination under 10 CFR 50.59. An overall conclusion that no unreviewed safety l
question (USQ) is involved is acceptable so long as any changes in specific commitments in the J
l licensing basis have been thoroughly evaluated in reaching the overall conclusion. If the overall cumulative assessment leads a licensee to conclude a USQ is involved, incorporation of the GlP-2 methodology into the licensing basis would require the licensee to seek an amendment
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. under the provisions of 10 CFR 50.90.
l Principal Contributors: P. Y. Chen D.Jong Date:
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a REFERENCES
- 1. " Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Power Plant Equipment," Revision 2, corrected Feoruary 14,1992, Seismic Qualification Utility Group.
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- 2. NRC " Supplement No.1 to Generic Letter 87-02 including Supplemental Safety Evaluation Report No. 2 on Seismic Qualification Utility Group's Generic implementation Procedure, Revision 2, corrected February 14,1992," dated May 22,1992.
- 3. Letter, J. D. Woodard, SNC to USNRC, " Evaluation of the Joseph M. Farley Nuclear Plant, Unit 1,120-Day Response to Supplement 1 to Generic Letter 87-02," dated September 10, 1992.
- 4. Letter, NRC to SNC, " Evaluation of the Joseph M. Farley Nuclear Plant, Unit 1,120-Day Response to Supplement No.1 to Generic Letter 87-02," dated November 20,1992.
- 5. Letter with enclosures, Dave Morey, SNC, to NRC, Unresolved Safety issue A-46 Summary Report for Farley Nuclear Plant - Unit 1, " dated May 18,1995.
- 6. Letter, NRC to SNC, " Request for Additional Information on USl A-46-Joseph M. Farley Nuclear Plant, Unit 1," dated September 9,1995.
- 7. Letter, NRC, to SNC, " Request for Additbnal Information, Joseph M. Farley Nuclear Plant, Unit 1," dated August 29,1997.
- 8. Letter, NRC to SNC, " Request for AdditionalInformation," dated May 15,1997.
- 9. Letter, with attachments, Dave Morey, SNC, to NRC, " Response to the NRC's USl A-46 Request for Additional information for Farley Nuclear Plant - Unit 1," dated October 11, 1995.
- 10. Letter, with enclosure, Dave Morey, SNC, to NRC " Response to NRC USl A-46 Request for Additional information for Farley Nuclear Plant-Unit 1," dated October 28,1996.
- 11. Letter, with attachments, Dave Morey, SNC, to NRC, " Response to the NRC's USl A-46 Request for Additional Information for Farley Nuclear Plant - Unit 1," dated August 11,
- 1997,
- 12. Kennedy, R.P.M., et al.," Review Procedures to Assess Seismic Ruggedness of Can!ilever Bracket Cable Tray Supports," Revision 3.0, Senior Seismic Review and Advisory Panel, March 1,1991.
i?. ' Procedure for Evaluating Nuclear Power Plant Relay Seismic Functionality," Final Report, NP-7148-SL. Electric Power Research Institute, Palo Alto, Califomia, December 1990.
- 14. Letter, Dave Morey, SNC, to NRC Document Control Desk, " Notification of Completion, USl A-46, GL 87-02," December 20,1996.
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