ML17265A683

Safety Evaluation Re Response to Supplement 1 to GL 87-02 for Ginna Nuclear Power Plant on USI A-46

ML17265A683
Person / Time
Site:	Ginna
Issue date:	06/17/1999
From:	NRC (Affiliation Not Assigned)
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Shared Package
ML17265A682	List: ML17265A681 ML17265A683
References
REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR GL-87-02, GL-87-2, NUDOCS 9906220238
Download: ML17265A683 (26)

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UNITED STA>ES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATIONBY THE OFFICE OF NUCLEAR REACTOR REGULATION EVALUATIONOF ROCHESTER GAS AND ELECTRIC CORPORATION'S RESPONSE TO SUPPLEMENT NO. 1 TO GENERIC LETTER 87-02 GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244

1.0 BACKGROUND

In December 1980, the NRC designated "Seismic Qualification of Equipment in Operating Plants" as Unresolved Safety Issue (USI) A-46. The safety issue of concern was that equipment in nuclear plants for which construction permit applications had been docketed before about 1972 had not been reviewed according to the 1980-81 licensing criteria for the seismic qualification of equipment, such as Regulatory Guide (RG) 1.100 (Reference 1), IEEE Standard 344-1975 (Reference 2), and Section 3.10 of the Standard Review Plan (NUREG-0800, July 1981) (Reference 3). To address USI A-46, affected utilities formed the Seismic Qualification UtilityGroup (SQUG) in 1982.

The NRC staff issued Generic Letter (GL) 87-02 in February 1987 (Reference 4) to provide guidance for the resolution of USI A-46. It concluded that the seismic adequacy of certain equipment in operating nuclear power plants should be reviewed against seismic criteria not in use when these plants were being constructed.

In 1987, SQUG, representing its member utilities, committed to develop a Generic Implementation Procedure (GIP) for implementing the resolution of USI A-46. SQUG requested a deferment of the 60-day response period, as requested in GL 87-02, until after the NRC issues its final safety evaluation report (SER) on the final version of GIP.

On May 22, 1992, the staff issued Supplement No.1 to GL 87-02 which transmitted its final SER (SSER No. 2, Reference 5) on the then final version of GIP (GIP Revision 2, as corrected on February 14, 1992, or simply GIP-2, Reference 6). In the supplement to GL 87-02, the staff requested that USI A-46 licensees who are members of SQUG to either provide a commitment to use both the SQUG commitments and the implementation guidance described in GIP-2, as supplemented by the staff's SSER No. 2, or else provide an alternative method for responding to GL 87-02.

In a letter dated September 21, 1992 (Reference 7), Rochester Gas and Electric Corporation (RG8 E), the licensee for the Ginna Nuclear Power Plant and a member of SQUG, committed to the implementation of GIP-2 for resolving USI A-46 at the Ginna plant. The NRC subsequently approved tlie licensee's approach and schedule in a letter dated November 30, 1992 (Reference 8).

'7906220238 O'F0617 PDR ADQCK 05000244 P

PDR Enclosure By letter dated January 31, 1997 (Reference 9), the licensee submitted a report summarizing the results of its USI A-46 implementation program.

The staff reviewed the report and issued a request for additional information (RAI) on April6, 1998 (Reference 10). The licensee subsequently submitted its response to the RAI in a letter dated May 27, 1998 (Reference 11).

The staff reviewed the licensee's response and determined that further information was required from the licensee in order for the staff to complete its review. A second RAI was issued to the licensee on December 3, 1998 (Reference 12), to which the licensee responded on February 2, 1999 (Reference 13). Teleconferences between NRC staff and RG&E were subsequently held on April26, 1999, and May 6, 1999.

In these teleconferences, the NRC staff discussed the appropriateness of the licensee's application of Method A.1 of GIP-2 in determining the seismic demand for equipment within 40 feet above the grade level. As a result, RG&E submitted, in a letter dated May 25, 1999 (Reference 14), additional information to support its use of Method A.1 of GIP-2 for equipment items located at various etevations of Ginna buildings.

This report provides the staff evaluation of the licensee's USI A-46 implementation program based on the staff's review of the summary report, supplemental information and clarification provided by the licensee in response to the staff's RAls.

2.0 DISCUSSION AND EVALUATION The staff's review of the Ginna USI A-46 summary report (Reference 9) consisted of a screening-level review of specific sections of the licensee's program, with emphasis placed on identification and resolution of outliers, i.e., equipment items which did not readily pass GIP-2 screening and evaluation criteria. The report identifies a safe shutdown equipment list (SSEL) and contains the screening verification and walkdown of mechanical and electrical equipment.

The summary report also contains relay evaluations and the evaluation of seismic adequacy for tanks and heat exchangers, cable and conduit raceways, and the identification and resolution of outliers, including the proposed resolution schedules.

2.1 Seismic Demand Determination Ground S ectra and In-structure Res onse S ectra Ginna Nuclear Power Plant was designed for an operating basis earthquake characterized by a peak horizontal ground acceleration of 0.08g and reviewed for a safe shutdown earthquake (SSE) with a peak horizontal ground acceleration of 0.20g. The peak horizontal and vertical accelerations were assumed to be the same.

The ground response spectra used were those developed by Housner.

Most Class I structures and equipment were analyzed by the equivalent static method (Reference 15).

Ginna was one of the nuclear power plants which participated in the systematic evaluation program (SEP).

During that program it was determined that the appropriate ground motion to be used to evaluate Ginna for the SEP was a site-specific spectrum with a peak ground acceleration of 0.17g.

RG&E chose to use a RG 1.60 spectrum with a peak ground acceleration of 0.20g to develop in-structure response spectra (IRS) for use in the SEP-and in several other programs which required in-structure response spectra.

In the120-day response to GL 87-02, Supplement 1 (Reference 7), the licensee stated, "Since Ginna is identified as a Category 2 (SEP) plant in GL 87-02, Supplement 1, RG&E will use the

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e options provided in GIP-2 for defining seismic demand (i.e., appropriate median-centered and conservative design IRS, depending on the building, the location of equipment in the building, and equipment characteristics)."

The NRC staff provided its evaluation of Reference 2 in a letter on November 30, 1992 (Reference 8). In response to questions raised by the NRC staff, in its evaluation, about which IRS RG&E planned to use in its A-46 program, RG&E provided additional information in Reference 16.

In Reference 16, RG&E stated, "There are several IRS contained within licensing basis documents for Ginna as a result of previous industry bulletins and programs.

As an alternative to performing an extensive review of these IRS, RG&E plans to use the spectra developed by the sbismic piping upgrade program (Engineering Work Request (EWR) No.

2512), which was initiated as a result of IE Bulletins 79-02 and 79-14, and the SEP. The seismic upgrade program evaluated safety-related piping against current criteria (e.g., RG 1.60, RG 1.61, SRP) and was reviewed and accepted by the NRC (References e and f)~ The IRS developed by EWR No. 2512 have been subsequently used at Ginna for plant modifications.

RG&E therefore maintains that the IRS developed by EWR No. 2512 meet the definition of "conservative design" spectra as provided in Section 4.2.4 of the GIP and associated SSER No.

2." The NRC staff has" found the characterization of the IRS as being conservative design spectra appropriate and the IRS are acceptable as the seismic demand for use in the A-46 program for Ginna.

2.2 Seismic Evaluation Personnel The screening verification, walkdown, and outlier identification was performed by a seismic

'review team (SRT), comprised of seismic capability engineers, as defined in GIP-2. GIP-2 describes the responsibilities and qualifications of the individuals who willimplement this generic procedure.

For a complete resolution of the USI A-46 issue, the seismic evaluation personnel should include individuals with sufficient expertise to identify safe shutdown equipment, perform the plant walkdown and verify the seismic adequacy of equipment and cable/conduit raceway systems, and to perform the relay screening and evaluation.

This involves a number of plant and engineering disciplines including structural, mechanical, electrical, system, earthquake, and plant operations.

Based on the information provided in Appendix A to the Seismic Evaluation Report (Reference 9) and Reference 11, the staff concludes that the qualifications of the individuals responsible for implementing the resolution of USI A-46, including the third party reviewers, meet the criteria of GIP-2 and the staff's SSER No. 2, and are, therefore, acceptable to perform the USI A-46 program for Ginna.

2.3 Safe Shutdown Path GL 87-02 specifies that the licensee should be able to bring the plant to a hot shutdown condition and maintain it in this state for 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 January 31, 1997, submittal (Reference 9), the licensee addressed the following plant safety functions: reactor reactivity control, pressure control, inventory control, and decay heat removal. A primary and an alternate safe shutdown success path with their support systems and instrumentation were identified for each of these safety functions to ensure that the plant is capable of being brought to, and maintained in, a safe 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. Attachment II to the seismic evaluation report (Reference 9) provides the SSEL.

The reactor decay heat removal function is accomplished by relieving steam from the reactor by automatic operation of main steam safety valves until such time that the decay heat rate decreases to the point where atmospheric relief valves (ARVs) can be used.

Makeup water to the steam generators willbe supplied by the auxiliary feedwater system which takes suction from the condensate storage tank. Once the condensate storage tank has been depleted after a few hours, the standby auxiliary feedwater system is realigned to the service water system to achieve long term decay heat removal and cooldown. Lake Ontario is the source of the water supply for the service water system.

The plant operations department reviewed the equipment listed in Attachment II against the plant operating procedures and concluded that the plant operating procedures and operator training were adequate to establish and maintain the plant in a safe shutdown condition following a postulated SSE.

The staff concludes that the licensee's approach to demonstrate that it can 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 /> following a seismic event is acceptable for use in the USI A-46 program at Ginna.

2.4 Seismic Screenin Verification and Walkdown of Mechanical and Electrical E ui ment The staff's evaluation focused primarily on the licensee's identification and resolution of equipment outliers, i.e., equipment items which do not comply with all of the screening guidelines provided in GIP-2. GIP-2 screening guidelines are intended to be used as a generic basis for evaluating the seismic adequacy of equipment.

However, if an item of equipment fails to pass these generic screens, it may still be shown to be adequate by additional evaluations.

2.4.1 E ui ment Seismic Ca acit Com ared to Seismic Demand In the "R. E Ginna USI A-46 Seismic Evaluation Report" attached to the licensee's January 31, 1997, submittal (Reference 9), R&GE stated that all the equipment on the seismic SSEL are housed in one of the following structures:

AuxiliaryBuilding (AB)

Standby AuxiliaryFeedwater Pump Building (AF)

Control Building (CB)

Diesel Generator Building (DG)

Intermediate Building (IB)

Reactor Containment (RC)

Screen House (SH)

Turbine Building (TB)

'h GIP-2 requires a comparison of the seismic capacity of.the electrical and mechanical equipment to the seismic demand.

The seismic capacity is based on the SQUG earthquake experience data base as represented by the bounding spectrum (BS), 1.5 times the BS (reference spectrum) or the generic seismic testing data as represented by the Generic Equipment Ruggedness Spectra (GERS). The seismic demand is represented by the plant's safe shutdown (SSE) ground response spectra (GRS) and IRS. RG&E did not use the GERS for the seismic capacity of equipment at Ginna. Since RG&E did not use GERS, there were two GIP-2 methods available for comparing equipment seismic capacity to the seismic demand.

Method A.1 compares the SQUG BS to the plant's safe shutdown GRS and Method B.1 compares 1.5 times the BS to the conservative design spectra or to the median-centered IRS.

GIP-2 places limitations on the use of Method A.1. These limitations are that the SSE GRS can be used for comparison to the BS when:

The equipment is mounted in the nuclear plant at an elevation below about 40-feet above the effective grade.

The equipment, including its supports, have a fundamental natural frequency greater than about 8 Hz.

The amplification factor between the free field GRS and the IRS is not more than about 1.5.

In Reference 9, the licensee indicates that:

The auxiliary building floor response spectra at elevations 253 and 271 feet and reactor containment floor response spectra at elevations 245, 253, 267 and 278 feet are largely enveloped by 1.5 times the BS. The floor response spectra exceed 1.5 times the BS at higher frequencies (about 20 Hz), but these exceedances are not very large (note that the floor response spectra are 4 percent of critical damping).

Equipment at these elevations meet the capacity versus demand requirement irrespective of their fundamental frequency.

The CB floor response spectra at elevations 253, 271 and 289 feet, DG building floor response spectra at elevations 253 and 285 feet, IB floor response spectra at elevations 253 and 271 feet, and the SH floor response spectra at elevation 253 feet all exceed 1.5 times the bounding spectrum.

However, these elevations are all within 40 feet of grade for the respective buildings, so the equipment at these elevations meet the capacity versus demand requirement as long as their fundamental frequency is greater than 8 Hz.

The standby AF pump building floor response spectra are enveloped by 1.5 x the BS at frequencies greater than about 6 Hz. Section 4.2 of the GIP states "The seismic capacity spectrum needs only to envelop the seismic demand spectrum for frequencies at and above the conservatively estimated lowest natural frequency of the item of equipment being evaluated." Therefore, equipment in the standby AF pump building meet the capacity versus demand requirements as long as their fundamental frequency is greater than about 6 Hz.

The seismic capacity versus demand basis used for each item of equipment on the SSEL are provided in the screening verification data sheets (SVDS) in Appendix B of Reference 9. There are a total of 236 equipment items listed in the SVDS. Method A.1 (BS versus GRS) was used for 207 of these items; Method B.1 (1.5 BS versus. IRS) was used for 15 of these items; and for the remaining 14 items, the design of the component was compared to a realistic response spectrum.

The NRC staff performed a comparison of the GRS to the IRS for each of the elevations below about 40 feet above the building's effective grade to determine if the amplification limitof 1.5 is exceeded.

The GRS used is the 5 percent of critical damping RG 1.60 spectrum with a peak ground acceleration of 0.2 g. The IRS used are the spectra RG&E sent to the NRC by telephone facsimile on August 5, 1998. The labels on the diagrams of these IRS indicate that they are for 4 percent of critical damping except for those at elevations 253 feet and 271 feet in the IB which indicate that they are for 5 percent of critical damping.

The staff has found, that for these IRS, the 4 percent of critical damping spectra are about 10-percent higher in amplitude than the 5 percent of critical damping spectra at frequencies above 8 Hz. Thus, the ratio of the 4 percent of critical damping IRS to the 5 percent of critical damping GRS has a 10 percent overestimate.

For the standby AF pump building, RG&E states that the grade elevation is 271 feet (Reference 13). A comparison of the GRS to the IRS at elevations 296 feet and 269 feet shows that the amplification factors at both levels are about 1.5 or less for frequencies above about 8 Hz.

For the reactor containment, RG&E states that the grade elevation is 235 feet (Reference 13).

A comparison of the GRS to the IRS at elevations 245 feet and 253 feet shows that the amplification factors at both levels are about 1.5 or less for frequencies above about 8 Hz. The amplification factor at elevation 267 feet is about 1.5 for frequencies between 8 and 16 Hz and the amplification factor ranges from about 2 to about 2.3 for frequencies above 16 Hz. The amplification factor at elevation 278 feet is about 1.5 for frequencies between 8 and 15 Hz and the amplification factor ranges from about 2.6 to about 3.1 for frequencies above 15 Hz.

For the AB, RG&E states that the grade elevation is 271 feet (Reference 13). A comparison of the GRS to the IRS at elevation 253 feet shows that the amplification factor is between 1.5 and 2 for frequencies above 8 Hz. The amplification factor at elevation 271 feet is between 1.8 and 2 for frequencies between 8 and 12 Hz and the amplification factor ranges from about 2.1 to about 2.6 for frequencies above 12 Hz.

For the SH, RG&E states that the grade elevation is 253 feet (Reference 13). A comparison of the GRS to the IRS at elevation 253 feet shows that the amplification factor is about 1.5 for frequencies between about 8 and 10 Hz. For frequencies above 10 Hz the amplification ranges from about 2.3 to 3.3.

For the DG building, RG&E states that the grade elevation is 253 feet (Reference 13). A comparison of the GRS to the IRS at elevation 253 feet shows that the amplification factor above 8 Hz is from about 1.8 to 4.5 depending on the frequency.

The amplification factor at elevation 285 feet is between 2.6 and 8.7 for frequencies of 8 Hz and above depending on the frequency.

For the CB, RG&E states that the grade elevation is 271 feet (Reference 13). A comparison of the GRS to the IRS at elevation 253 feet shows that the amplification factor at frequencies above 8 Hz has a range from 1.8 to 3.4 depending on the frequency.

The amplification factor at elevation 271 feet is between 1.8 and 3.4 for frequencies of 8 Hz and above depending on the frequency.

The amplification factor at elevation 289 feet is between 2.2 and 7.6 for frequencies of 8 Hz and above depending on the frequency.

For the IB, RG8 E states that the grade elevation is 253 feet (Reference 13). A comparison of the GRS to the IRS at elevation 253 feet shows that the amplification factor is between 2.5 and 3.9 at frequencies between 10 and 20 Hz depending on the frequency.

The amplification factor at elevation 271 feet is between 2.5 and 3.7 for frequencies of 8 Hz and above depending on the frequency.

For the TB, RG&E states that the grade elevation is 253 feet (Reference 13). According to the SVDS in Reference 9, there is only one SSEL equipment item in the TB, the TB service water isolation valve MOV-4613, at elevation 253 feet.

RG8 E did not provide an IRS for the TB in any of its A-46 submittals nor is there an IRS for this structure in the Updated Final Safety Evaluation Report ('FSAR) ~

In Reference 13, RG&E presented a position paper on the use of Method A.1 at Ginna.

It states that its purpose is to describe many of the conservatisms that exist in computed IRS.

Among the qualitative assessments of the conservatisms in the Ginna IRS, RG&E presented the following arguments:

Most computed in-structure spectra have very narrow, highly amplified peaks at the resonant frequency of the structure.

In most cases these narrow, highly amplified peaks are artificiallybroadened to account for uncertainty in the structure's natural frequency.

2.

The design earthquakes (OBE and SSE) were defined at the soil surface.

However, since the IRS for Ginna were generated using a model defining the input motion at the foundation level, conservatism exits in the IRS.

3.

The IRS were generated for 4 percent of critical damping instead of the 5 percent of critical damping specified for the A-46 program.

4.

The RG 1.60 spectral shape used as the ground motion input for generation of the IRS is broader and has higher spectral amplifications than the design basis Housner spectrum The Ginna analyses did not include any reduction in foundation motion due to embedment effects, wave scattering effects and radiation of energy from the structure into the surrounding media which would be obtained if a soil structure interaction analysis were part of the structural analyses.

6.

Actual measured in-structure response spectra on "nuclear type" structures support the 1.5 response levels advocated within Method A.1 of GIP-2.

7.

The development did not include soil structure interaction which would have reduced the

-amplification of the IRS due to soil damping and wave scattering effects.

The NRC staff reviewed the qualitative assessments provided by RG&E in our evaluation of the use of Method A.1. The staff found some of these points germane and took them into account; but some are redundant and some are not relevant.

For example, the actual measured IRS at Humboldt Bay (1975 Ferndale earthquake) and the Pleasant Valley Pumping Station (1983 Coalinga earthquake), which are cases quoted by RG&E in Reference 13; demonstrate the effect of embedment rather than reduced amplitudes at higher elevations in a structure as is claimed in Reference

13. This is obvious because the in-structure instruments are at elevations 60 or more feet lower in elevation than are the ground surface instruments.

With respect to Item 4 above, the staff calculated the amplification factor comparing the input ground motion (RG 1.60 spectrum) to the IRS. This is the appropriate comparison.

With respect to Item 7 above, the use of soil structure interaction analysis generally does not result in significant reductions of the IRS for structures founded on rock, lean concrete or engineered compact fill as are the structures at Ginna.

The staff carefully considered the qualitative arguments made by RG8 E to justify the use of GIP-2 Method A.1 to compare the seismic demand to seismic capacity at locations below 40 feet above grade elevation and concluded that there were some locations where the use of GIP-2 Method A.1 for A-46 at Ginna had not been adequately justified by these qualitative arguments.

These locations were the CB elevation 289 feet, the DG building elevation 253 feet and elevation 285 feet, and the IB elevation 253 feet and elevation 271 feet if the equipment has a fundamental natural frequency in the 8 to 10 Hz frequency range.

Teleconferences between NRC staff and RG&E were held on April26, 1999, and May 6, 1999.

In these teleconferences, NRC staff explained that quantitative information was needed to evaluate the amount of conservatism in the design in-structure response for the Ginna buildings. As a result, RGLE submitted additional information (Reference 14) to further justify the use of Method A.1 at locations where the amplification factor is larger than about 1.5. Reference 14 contains a comparison evaluation of overall margins between median-centered analysis and design-basis analysis for nuclear power plant structures at other facilities similar in construction, building frequency and damping to those at the R. E. Ginna Nuclear Power Plant.

The median-centered spectra and the conservative design spectra for five (5) reinforced concrete buildings at four (4) nuclear power plants were presented in Reference 14. The reinforced concrete shear wall structures are two (2) ABs, a RB interior structure, a RB exterior shell, and a containment interior structure which are typical of those found at nuclear power plants. The ratios of the conservative design spectra to median-centered spectra were 2.53, 5.3, 3.3, 2.3, and 5.4. The staff considers the wide range of the ratios of the conservative design spectra to median center spectra to be due to the different methods and levels of conservatism used in the analyses of the structures rather than differences between structural configurations.

The mean of the ratios is 3.77. The staff used this mean value to estimate the amplification factor for the Ginna structures.

If median-centered spectra were developed for locations in Ginna where Method A.1 was used, the staff estimates that such an amplification factor would be about 1.5.

The staff considers the use of the above approximation for estimating the level of the median-centered IRS to be adequate to justify the use of GIP-2 Method A.1 for equipment in the A-46 SSEL. The staff's acceptance of this approximation was based primarily on the fact that the floor spectra for the building structures at Ginna are extremely conservative.

The use of this approach is limited to the verification of seismic adequacy of mechanical and electrical equipment at Ginna..The use of this'approach for any other application will require staff review and approval.

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2.4.2 Assessment of E ui ment Caveats As a second screening guideline, the licensee verified the seismic adequacy of an item of mechanical or electrical equipment by confirming that (1) the equipment characteristics are generally similar to the earthquake experience equipment class or the generic seismic testing equipment class, and (2) the equipment meets the intent of the specific caveats for the equipment class.

This review is necessary only when the BS, 1.5 times the BS, or the GERS are used to represent the seismic capacity of equipment.

If equipment-specific seismic qualification data is used instead, then only the specific restrictions applicable to that equipment-specific qualification data need be applied.

The "caveats" are defined as a set of inclusion and exclusion rules, which are established to represent specific characteristics and features particularly important for seismic adequacy of a particular class of equipment.

Appendix B of GIP-2 contains a summary of the'caveats for the earthquake experience equipment class and for the generic seismic testing equipment class.

Another aspect of verifying the seismic adequacy of equipment included within the scope of this procedure is explained by the "rule of the box."

For the equipment included in either the earthquake experience or testing equipment class, all of the components mounted on or inside this equipment are considered to be part of that equipment and do not have to be evaluated separately.

In Reference 9, the licensee identified a total of 234 items of mechanical and electrical equipment on the seismic SSEL, excluding equipment designated as "rule-of-the-box" type equipment, or tanks.

The licensee provided a brief description of the 20 classes of equipment on the seismic SSEL, including the number of items in the equipment class, the location of the equipment, and the manufacturer.

A number of these equipment items were found to have their characteristics outside the bound of the above defined caveats.

These items are identified in Table 3.2 of Reference 9 as outliers, together with the licensee's recommended resolution.

The licensee's proposed approaches for resolution we"e found to be generally acceptable to the staff. In the licensee's A-46 program, there were no instances in which the intent of a caveat is met without meeting the specific wording of the caveat.

2.4.3 E ui ment Anchora es As a third screening guideline, the licensee described in Section 3.1.3 of the seismic evaluation report the procedure used for verifying the seismic adequacy of equipment anchorages.

The licensee discussed equipment anchorages used at Ginna by anchorage type, and described the type of equipment anchored in that manner and how it was evaluated.

Those items of equipment that did not meet GIP-2 anchorage requirements were identified as outliers.

The licensee stated that bolt tightness checks, field sketches, and calculations were performed for anchorages, and they were documented in the SEWS.

Since the concrete compressive

'trength was specified as 3,000 psi for Ginna, the anchor strength reduction factor was applied in accordance with Appendix B of GIP-2.

There are three types of anchorages at Ginna: cast-in-place bolts, concrete expansion bolts, and welding to embedded steel.

Allof the horizontal and vertical pumps, the diesel generators, and the DGs'xhaust mufflers are anchored with cast-in-place bolts. A few items of floor-mounted electrical equipment are anchored with through bolts and they were evaluated as cast-in-place bolts. Anchorage details and equipment weights were obtained from plant drawings, and anchorage calculations were performed per the requirements of GIP-2 and Appendix B.

Almost all of the electrical equipment (motor control centers (MCCs), switchgear, transformers, distribution panels, battery racks, chargers/inverters, and cabinets), the DG room cooling fans, and the mechanical instruments (Class 18) are anchored with concrete expansion anchors.

Anchorage evaluations were performed per the requirements of GIP-2 and its Appendix C. Bolt tightness was verified by one of three ways. Small wall-mounted equipment, such as distribution panels, were tug-tested by the SRT.

In most cases, this is the only practical option because the expansion anchors for equipment of this type are behind the panel and not accessible.

For anchors found, as a result of the SEP program, to be required to anchor the electrical equipment, bolt tightness checks were not repeated during the A-46 evaluation because those anchors were installed using a QA/QC controlled procedures considered by the SRT to be more stringent than the GIP-2 requirements.

During the SQUG SSEL equipment walkdown effort, anchorage installation and SEP anchorage torque data sheets were reviewed.

The bolt installation torque values were found to be higher than GIP-2 inspection requirements.

The bolt tightness checks for all other anchors were performed by the SRT using a hand wrench. The staff considers the licensee's methods adequate for the A-46 program.

Welding to embedded steel was used to suspend the air handling units (AHUs) from the ceiling in the standby AF pump building. The AHUs are bolted to a steel frame which is welded to steel embedded in the ceiling. The embedded steel is anchored with Nelson studs in.the concrete.

Anchorage details and the AHUs weights were obtained from plant drawings.

Detailed anchorage calculations were performed per the requirements of GIP-2, Section 4.4 and Appendix B. The calculations considered both the strength of the welds and the embedded steel's anchorage.

The staff considers the licensee's evaluation adequate for the A-46 program.

The staff finds that the licensee, by following GIP-2 procedures, has established a viable program to verify the adequacy of equipment anchorages.

Equipment anchorages that could not pass the above screening criteria were identified as outliers in Table 3.2 of the seismic evaluation report. For example, Westinghouse motor control centers, designated as MCC-L and -M, are anchored with eight (8) ~/a" expansion anchors into a 1.5" to 2" thick concrete pad, where a number of the anchors have significant exposed length, so their exact anchorage embedments were not certain.

For their resolution, the SRT recommended that the embedment be determined via UT test, a bolt tightness check be performed, and then the anchorage evaluation attached to the SEWS be adjusted accordingly. Other anchorage issues include the required upgrade of the anchorage for the Battery Racks A and B, in order to meet the GIP-2 requirements for side-to-side loads. The licensee had committed to complete all outlier resolutions prior to the plant startup from its 1999 refueling outage.

In a teleconference held on May 25, 1999, the licensee confirmed that all the Ginna A-46-related outliers have been resolved.

Specifically, the Westinghouse motor control centers, MCC-L and -M, have been evaluated and found acceptable without modification. New battery racks, on the other hand, have been installed for battery racks A and B. The staff finds the licensee's equipment anchorage evaluation acceptable for the resolution of USI A-46 at Ginna.

2.4.4 Seismic S atial Interaction Evaluation The licensee addressed potential spatial interaction effects for the equipment in Section 3.1.4 of the seismic evaluation report. This serves to ensure that there is no adverse seismic spatial interaction between the equipment under consideration and nearby equipment, systems, and structures which could cause the equipment to fail to perform its intended safe shutdown function during an SSE. The interactions of concern are (1) proximity effects, (2) structural failure and falling, a;d (3) flexibilityof attached lines and cables.

Guidelines forjudging potential interaction effects, when verifying the seismic adequacy of equipment, are presented in GIP-2, Section 4.5, and its Appendix D. Those items of equipment that did not meet the GIP-2 interaction requirements were considered outliers and are identified in Table 3.2.

During the plant walkdowns at Ginna, the SRT identified a few interaction concerns, including masonry block walls, which are a common interaction issue.

At Ginna, there are masonry block walls in the auxiliary, control, and IBs. As stated in Section 3.1.4 of the seismic evaluation report, the CB block walls have been evaluated and found adequate.

Some critical equipment in other areas have been protected using barriers.

Seismic SSEL equipment outside the CB that has not been protected and is near a block wall was identified as an outlier. The staff has reviewed the licensee's proposed resolutions for these outliers, as provided in Table 3.2, and found them to be reasonable.

2.5 Tanks and Heat Exchan ers A total of twelve tanks and heat exchangers were evaluated and the summary of the evaluations was documented in Table 5.1 of the seismic evaluation report. The detailed evaluations, including field notes, sketches, photographs, and calculations are documented in the SEWS. The evaluation results indicated that all tanks and heat exchangers have met the GIP-2 requirements except the refueling water storage tank (RWST), which was identified as an outlier. The RWST is a flat bottom steel vertical tank, 26.5 feet in diameter, 81 feet tall, and anchored at the bottom with thirty (30) 2.25" diameter cast-in-place bolts. The tank is founded at elevation 235-feet inside the AB, and penetrates the AB floor slabs at elevations 253 feet and 271 feet. The licensee stated that the demand of the bending moment at the bottom of the RWST tank exceeded the capacity by a significant margin, primarily due to the low elephant's foot buckling stress at the base of the tank. As indicated in Table 5.2, the licensee has designed and installed a lateral bracing, which attaches the tank to the floor slab of the AB at elevation 271 feet. The licensee has provided the design calculations to demonstrate the adequacy of such modification. The staff has reviewed the calculations and found the modification of the RWST to be acceptable.

The licensee's evaluation of the tank and heat exchangers is, therefore, adequate for the resolution of A-46 at Ginna.

2.6 Cable and Conduit Racewa s

The licensee stated that the raceway review was performed as specified in GIP-2 Section 8.

Raceway systems were examined during a walkdown, checked against the inclusion rules and

~ ~ other seismic performance concerns as specified in Section 8.2 of the GIP-2, and examined for seismic spatial interactions with adjacent equipment and structures.

According to the licensee, the scope of the review was determined by studying the plant layout drawings and walking through the plant to determine how cabling is routed to the areas of the plant containing the SSEL equipment.

The walkdown of each plant area was documented on a plant area summary sheet (PASS). Table 6.1 of the seismic evaluation report summarizes the PASS and includes the supports selected for limited analytical reviews (LAR).

Eleven representative worst-case raceway supports were selected for LAR per GIP-2 Section 8.3. For the LAR reviews of the raceway supports, which are anchored with Q-deck inserts, the licensee first used a conservative allowable load of 2000 Ib for all the evaluations of dead load, vertical capacity check (3 times dead load), and lateral load. Allsuch supports passed the dead load evaluations, but several supports exceeded the 2000 Ib allowable for the vertical capacity check and/or lateral load evaluations.

These hanger supports were identified as outliers. As indicated in Table 6.2 of the seismic evaluation report, however, these supports were later found to have a demonstrated capacity of at least 5700 Ib, and were considered acceptable.

Also, as indicated in Table 6.2, Hanger ¹ AB34 failed the LAR dead load evaluation due to excessive moment developed in the stringer. This hanger was identified as an outlier. In Reference 14, the licensee has confirmed that the hanger has been modified to meet the GIP-2 requirements.

This is acceptable to the staff for the A-46 program.

A number of raceway hangers are attached or adjacent to masonry block walls with unknown seismic capacity. These supports were identified to be outliers, pending the demonstration of the seismic adequacy of the associated block walls. The staff reviewed the descriptions of the outliers and their resolutions or proposed resolutions and determined them to be reasonable..ln Reference 14, the licensee further confirmed that all the cable tray and conduit raceway outliers have been resolved to meet the GIP-2 guidelines.

This is acceptable to the staff for the resolution of A-46 program at Ginna.

A review of relays associated with safe shutdown equipment is required as part of the resolution of USI A-46 program.

The purpose of the relay review is to verify that safe shutdown systems would not be prevented from performing their safe shutdown functions because of relay (contact) chatter during the period of strong ground motion associated with an SSE. A total of 1357 relays (71 essential and 1286 non-essential) were identified to be associated with components on the relay review SSEL. The methodology utilized for relay evaluation at Ginna was similar to that described in GIP-2.

2.7.1 Identification of Essential Rela s The licensee's relay evaluation report (Reference 9) indicated that the SRT relay review was conducted utilizing the following process:

(1)

A SSEL of equipment that must be evaluated for a relay review was developed.

This list of equipment is based on (i) active electrically controlled or powered safe shutdown equipment whose function could be adversely affected by relay malfunction and (ii)

I inactive safe shutdown equipment for which relay malfunction could cause unacceptable spurious operation.

(2)

Plant electrical drawings of the circuits associated with the above safe shutdown equipment were used to identify relays that need to be evaluated for potential malfunction. The following assumptions were utilized in establishing the scope of the relay review:

a)

Relay willnot be damaged by an earthquake, with the exception of fragile relays.

b)

Unqualified relays are assumed to chatter during the short period of strong motion during an earthquake.

c)

Solid-state relays and mechanically actuated switches are considered to be seismically rugged and need not be evaluated for contact chatter.

(3)

Relays whose malfunction willnot prevent the achievement of any safe shutdown function and willnot cause unacceptable spurious operation of equipment are designated as "non-essential" and do not need further review. The seismic adequacy of the remaining, "essential" relays is required to be assessed.

2.7.2 Seismic Screenin of Essential Rela s The screening procedures are described in GIP-2, Section 6. Of the 65 relays identified as essential, utilizing the above procedure, 53 were found acceptable and 12 did not pass the screening criteria. Of the 53 essential relays found acceptable, 24 were screened using screening level 1, 19 were screened using screening level 2, and 10 were screened because they are not operating norma! Iy in the open (NOPNO) state.

Screening level 1 requires that the SSE GRS be enveloped by the BS, the cabinet housing the relay be mounted within about 40 feet above the effective grade elevation, and the relay not be one of the known "low-ruggedness" types and have a capacity of at least 8g. Cabinet locations and relay capacities are summarized in Table 4.3 and Table 4.2 of the seismic evaluation report, respectively.

Screening Level 2 requires that the relay capacity be greater than the demand at the base of the cabinet multiplied by an amplification factor. Allof the 19 essential relays screened in this manner are Westinghouse A200 motor starters.

These relays have a capacity of 4.5g, and are housed in MCCs, which have an amplification factor of 3. The demand at the base of the MCC cabinets is indicated in Table 4.3, as ranging between 0.82g to 0.88g. Therefore, all these relays are acceptable.

The remaining 10 relays that screened had a low capacity when required to operate in the not operating normally closed (NOPNC) state, and a high capacity otherwise.

These relays are in the NOPNO state, and therefore they screen.

Of the 12 essential relays that did not screen, 10 lack relay capacity data and 2 have a capacity less than the demand computed per GIP-2 Screening level 2. Allof these relays are designated outliers, as indicated in Table 4.1 of the seismic evaluation report. They are also summarized in Appendix H of the seismic evaluation report, together with their resolution plan and schedule.

The staff has reviewed the above resolution plan and found them to be reasonable.

In Reference 14, the licensee further confirmed that all the Ginna A-46 related.

relay outliers have been completely resolved.

This is acceptable to the staff for resolution of A-46 program at Ginna.

2.8 Human Factors As ect As part of the resolution to USI A-46, the SQUG developed GIP-2 for use in part by licensees to identify and verify a safe shutdown equipment list (SSEL) and ensure adequate procedures and training were in place for plant operators to mitigate the consequences of an SSE.

GIP-2 described the use of operator action as a means of accomplishing those activities required to achieve safe shutdown. Section 3.2.7, "Operator Action Permitted," states, in part, that timely operator action is permitted as a means of achieving and maintaining a safe shutdown condition provided procedures are available and the operators are trained in their use.

Additionally, Section 3.2.6, "Single Equipment Failure," states that manual operator action of equipment which is normally power operated is permitted as a backup operation provided that sufficient manpower, time, and procedures are available.

Section 3.2.8, "Procedures,"

states, in part, that procedures should be in place for operating the selected equipment for safe shutdown and operators should be trained in their use.

It is not necessary to develop new procedures specifically for compliance with the USI A-46 program.

In Section 3.7, "Operations Department Review of SSEL," of GIP-2, the SQUG also described three methods for accomplishing the operations department reviews of the SSEL against the plant operating procedures.

Licensees were to decide which method or combination of methods were to be used for their plant-specific reviews. These methods included 1 ~ A "desk-top" review of applicable normal and emergency operating procedures,

2. use of a simulator to model the expected transient,

3. performing a limited control room and local in-plant walk-down of actions required by plant procedures.

The staff's evaluation of the SQUG approach for the identification and evaluation of the SSEL, including the use of operator actions, was provided in Section 11.3 of the staff's SSER on GIP-2. The evaluation concluded that the SQUG approach was acceptable.

The staff's review focused on verifying that the licensee had used on'e 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 actions could be used to achieve and maintain safe shutdown (e.g., resetting relays, manual operation of plant equipment).

The licensee provided information which outlined the use of the."desk-top" and "walk-down" evaluation methods by the Operations Department 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 licensee determined that the systems and equipment selected for seismic review in the USI A-46 program are those for which Normal, Abnormal, and Emergency Operating Procedures are available to bring the plant from a normal operating mode to a safe shutdown condition. The shutdown paths selected were reviewed by the R.E. Ginna Operations staff and determined that the procedures would provide adequate guidance to the operators in response to a seismic

C'

~ ~ event.

The licensee provided assurance that ample time existed for operators to take the required actions to safely shut down the plant. This had been accomplished during validation of the pertinent plant operating procedures related to the licensees A-46 program review.

The staff verified that the licensee had considered its operator training programs and verified that its training was sufficient to ensure that those actions specified in the procedures could be accomplished by the operating crews. The Operations Department verified that all actions necessary to safely shut down the plant were included in existing Normal, Abnormal', and Emergency Operating Procedures.

The licensee verified that no additional operator actions, beyond those associated with the safe shutdown paths, must be performed to bring the plant from a normal operating mode to a safe shutdown condition. As part of the licensee's validation of the SSEL paths, procedural enhancements primarily to the Earthquake Emergency Plan were implemented to provide operators additional reminders of expected worst case equipment losses and references to the procedure and step to remedy each of those potential losses.

In addition, the staff requested (Reference

19) verification that the licensee had adequately evaluated potential challenges to operators, such as lost or diminished lighting, harsh environmental conditions, potential for damaged equipment interfering with the operators tasks, and the potential for placing an operator in unfamiliar or inhospitable surroundings.

The licensee provided information (Reference 20) to substantiate that potential challenges to the operator were explicitly reviewed as part of the A-46 validation effort. The review determined that there were no newly created operator actions necessary to implement the SSEL paths selected.

However the licensee's review revealed that the primary obstacle to operators achieving safe shutdown was due to both direct and secondary damage incurred from intermediate building block wall failures. As a result, the licensee implemented an engineering modification to the affected wall panels which willpreclude seismically-induced failure and remove the potential for operational obstacles.

In addition, the licensee explicitly evaluated the potential for local failure of architectural features and the potential for adverse spacial interactions in the vicinityof safe shutdown equipment, where local operator action may be required, as part of the GIP-2 process (Reference 20). As a result of the review, several potential control room interaction sources were identified associated with non restrained equipment (e.g., an unsecured cabinet, copier, step ladder, and air sampler, unachored mail box, various stora'ge cabinets, Control Room ceiling tiles, and adjacent masonry walls). The licensee's Seismic Review Team directed removal of the identified seismic hazards, and verified that the Control Room ceiling tiles and masonry walls were evaluated and determined to be acceptable.

The licensee performed seismic interaction reviews which eliminated any concerns with the plant components and structures located in the immediate vicinityof the components which had to be manipulated.

Therefore the potential for physical barriers resulting from equipment or structural earthquake damage which could inhibit operator ability to access plant equipment was considered, and eliminated as a potential barrier to successful operator performance.

The licensee has provided the staff with sufficient information to demonstrate conformance with the NRC-approved review methodology outlined in the GIP-2 and is, therefore, acceptable.

2.9 Outiier Identification and Resolution As stated previously, an outlier is defined as an item of equipment which does not meet GIP-2 screening guidelines.

An outlier may be shown to be adequate for seismic loadings, by performing an additional evaluation using alternate methods or seismic qualification techniques acceptable to the staff. Based on the screening criteria stated in Section 2.4, a number of equipment items were identified during the walkdowns by the SRT as outliers. The licensee indicated in Section 3.3 of the seismic evaluation report that a total of twenty-seven (27) items of mechanical and electrical equipment, excluding tanks and heat exchangers as well as cable and conduit raceway systems, were designated as outliers. The staff has reviewed the licensee's proposed resolution of these outliers and found them to be reasonable.

As previously stated, of the twelve (12) tanks and heat exchangers that were evaluated as a part of the Ginna A-46 program, the RWST was identified as an outlier, due to inadequate bending moment capacity at the bottom of the tank. The licensee has resolved the outlier by installing a lateral bracing, which attaches the tank to the floor penetration of the AB at elevation 271 feet. The staff finds it to be acceptable.

Eleven representative worst-case raceway supports were selected for LAR per GIP-2 Section 8.3, and one was identified as outlier. This and the other raceway outliers associated with the PASS were identified in Table 6.3 of the seismic evaluation report. The staff has reviewed the descriptions of the outliers and their proposed resolutions, and found them to be reasonable.

As stated previously, for a total of 65 essential relays which were identified, 53 were found acceptable and 12 did not pass the screening criteria either due to lack of capacity data or for having a capacity less than the demand.

These 12 relays were designated as outliers. The staff has reviewed the licensee's proposed resolutions for these relay outiiers as provided in Appendix H to the seismic evaluation report and found them to be reasonable.

The licensee has indicated in Appendix H that all the modifications, including outlier resolutions, would be completed prior to returning to power from the 1999 refueling outage'.

In Reference 14, the licensee confirmed that outlier resolutions have been completed.

The staff has also reviewed an independent audit report by Dr. Robert P. Kennedy on the Ginna A-46 program, which is attached to the licensee's January 31, 1997, submittal (Reference 9).

In its RAI (Reference 10), the staff requested the licensee to address the comments and findings identified in the audit report. Specifically, the auditor commented on (i) the inconsistency in caveat status documented on the SVDS and SEWS, (ii) the use of IEEE-344 qualification for equipment as a basis for the adequacy of in-plant anchorage, (iii)the traceability of resolution for Bus 14 issues on load transfer between cubicles and lateral restraint of breakers, and (iv) the validity of IEEE-344 qualification as the basis for seismic capacity.

In its response provided in Reference 11, the licensee stated that the above audit was based on a draft of the seismic evaluation report. The draft report contained evaluation for the equipment that were separate from the SEWS and, in most cases, performed after the SEWS were completed.

This led to a number of review comments concerning the fact that some caveats noted as "U" (unknown) on the SEWS were listed as 'Y" (meaning acceptable) on the SVDS without clear traceability as to the justification. The licensee stated that prior to completion of the Ginna A-46 program, the SEWS would be re-worked and the seismic evaluation report reorganized so that the caveat status documented on the SVDS are identical to those on the SEWS. The licensee further stated that any arguments or calculations made to justify a caveat would either be included or specifically referenced in the SEWS. The licensee confirmed in Reference 14 that the above administrative work has been completed.

This is acceptable to the staff.

The licensee stated that the approach of utilizing the fact that an item of equipment had been IEEE-344 qualified as justification for accepting the anchorage was not used for the final report.

Instead, a specific anchorage evaluation based on GIP-2 requirements was performed for all equipment and is documented or referenced on the SEWS. This is acceptable to the staff.

In regard to the traceability of resolution for the identified Bus 14 issues on load transfer between cubicles and lateral restraint of breakers, the licensee stated in Reference 11 that the required traceability have been provided on the final SEWS for Bus 14. This is acceptable to the staff.

In regard to the validity of the use of IEEE-344 qualification as the basis for equipment seismic capacity, the licensee stated that the final SEWS contain a reference to the specific test document, the edition of IEEE-344 used, and a comment as to whether the test response spectrum envelopes the floor response spectrum used for the A-46 evaluations.

This is acceptable to the staff.

The audit report also indicated that the seismic margin assessment (SMA) methodology, described in the EPRI report, NP-6041 (Reference 17), has been used to evaluate the seismic adequacy of the reactor water make-up tank (RWMT), the Battery Racks (RTRYA& RTRYB),

and the battery room block wall. The staff indicated in its RAI of April6, 1998 (Reference 10),

that the methodology may yield analytical results which are not as conservative as those which could be obtained by following GIP-2 guidelines.

Because of the uncertainty of its conservatism, the methodology has not been endorsed by the staff for the analysis of safety-related systems and components, including the resolution of mechanical, electrical, and structural component outliers in the USI A-46 program.

The licensee stated in its May 27, 1998 submittal (Reference 11) that the RWMT is not on the SSEL submitted January 31, 1997 (Reference 9), and, therefore, is not within the scope of the Ginna A-46 program.

As for the battery racks, which are listed as SQUG outliers in Appendix H of the seismic evaluation report, the licensee stated that the battery racks would be modified during the 1999 refueling outage.

The licensee confirnied in Reference 14 that this has been completed.

The licensee also stated that the battery room block wall has been previously qualified for earthquake loading which was accepted by an NRC safety evaluation dated December 12, 1986. This wall qualification was based on use of nonlinear analysis supported by the San Onofre masonry wall test program. The licensee stated that the above equipment qualification has not relied on the use of the SMA approach, described in EPRI report, NP-6041 (Reference 17), and, therefore, there is no need to reevaluate these equipment items in order to address the Ginna A-46 program. The staff finds the licensee's response acceptable.

Based on its review of the licensee's responses to address the auditor's observation and findings discussed above, and the confirmation by the licensee in Reference 14, which indicates that all the Ginna A-46 outliers have been resolved, the staff concludes that the licensee's corrective actions are acceptable 3.0

SUMMARY

OF MAJOR STAFF FINDINGS Based on its review of the licensee's USI A-46 implementation program, as provided for each area discussed above, the staff did not identify any significant programmatic deviation from GIP-2 regarding the walkdown and the seismic adequacy evaluations at Ginna.

4.0 CONCLUSION

S The licensee's USI A-46 program at Ginna was established in response to Supplement 1 to GL 87-02 through a 10 CFR 50.54(f) letter. The licensee conducted the USI A-46 implementation in accordance with GIP-2. The licensee's submittal on its implementation of the USI A-46 program indicated that of the safe shutdown equipment list a total of twenty-seven (27) items of mechanical and electrical equipment, excluding tanks and heat exchangers as well as cable and conduit raceway systems, were designated as outliers. Twelve (12) tanks and heat exchangers were evaluated, and only the RWST was identified as an outlier, due to inadequate bending moment capacity at the bottom of the tank. The licensee has resolved the outlier by installing a lateral bracing for the tank.

Of the eleven representative worst-case raceway supports which were selected for LAR, one was identified as outlier. Other raceway outliers associated with the PASS were also identified by the licensee.

The staff has found the proposed resolutions of the outliers to be reasonable.

The licensee subsequently confirmed in Reference 14 that these outliers have been resolved.

The licensee also indicated that of the 65 relays identified as essential relays, 53 were found acceptable and 12 did not pass the GIP-2 screening criteria. Of the 53 essential relays found acceptable, 24 were screened using screening level 1, and 19 were screened using screening level 2. The remaining 10 relays that screened had a low capacity for the NOPNC state, and a high capacity otherwise.

These relays are in the NOPNO state, and therefore they were found acceptable.

Of the 12 essential relays that did not pass the screening criteria, 10 lack relay capacity data and 2 have a capacity less than the demand computed per GIP-2 screening leve 2. Allof these relays are designated as outliers.

As stated in the above, the licensee has confirmed in Reference 14 that all the unresolved outliers have been resolved prior to the plant startup from its 1999 refueling outage, in accordance with its commitment.

The staff concludes that the licensee's USI A-46 implementation program has, in general, met the purpose and intent of the criteria in GIP-2 and the staff's SSER No. 2 for the resolution of USI A-46. The staff has determined that the licensee's corrective actions will result in safety enhancements which, in certain aspects, are beyond the original licensing basis.

As a result, the licensee's actions provide sufficient basis to close the USI A-46 review at the facility. The staff also concludes that the licensee's implementation program to resolve USI A-46 at the facilityhas adequately addressed the purpose of the 10 CFR 50.54(f) request.

Licensee activities related to the USI A-46 implementation may be subject to NRC inspection.

Regarding future use of GIP-2 in licensing activities, the licensee may revise its licensing basis in accordance with the guidance in Section I.2.3 of the staff's SSER No. 2 on SQUG/GIP-2, and the staff's letter to SQUG's Chairman, Mr. Neil Smith, on June 19, 1998 (Reference 18).

Where the plant has specific commitments in the licensing basis with respect to seismic qualification, these commitments should be carefully considered.

The overall cumulative effect of the incorporation of the GIP-2 methodology, considered as a whole, should be assessed in making a determination under 10 CFR 50.59. An overall conclusion that no unresolved safety question (USQ) is involved is acceptable so long as any changes in specific commitments in the licensing basis have been thoroughly evaluated in reaching the overall conclusion.

If the overall cumulative assessment leads a licensee to conclude that a USQ is involved, incorporation of the GIP-2 methodo! gy into the licensing basis would require the licensee to seek an amendment under the provisions of 10 CFR 50.90.

Principal Contributors: A. J. Lee J. Ma R. Rothman G. S. Gallentti R. M. Pelton Date: June 17, 1999

5.0 REFERENCES

Regulatory Guide 1 ~100, "Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants," Revision 2, 1987 2.

IEEE Standard 344-1975, "IEEE Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations," dated January 31, 1975 3.

NRC SRP (NUREG-0800), Section 3.10, "Seismic and Dynamic Qualification of Mechanical and Electrical Equipment," Revision 2, July 1981 NRC GL 87-02, "Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46," February 1987 5.

"Supplemental Safety Evaluation Report No. 2 on Seismic Qualification UtilityGroup's Generic Implementation Procedure, Revision 2, corrected February 14, 1992" 6.

"Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Power Plant Equipment," Revision 2, corrected February 14, 1992, Seismic Qualification UtilityGroup 7.

8.

Letter, RG&E to NRC Document Control Desk, "Response to Supplement 1 to Generic Letter 87-02, SQUG Resolution of USI A-46," dated September 21, 1992 Letter, NRC to RG&E, "Evaluation of the Ginna Plant 120-Day Response to Supplement No. 1 to Generic Letter 87-02," dated November 30, 1992 9

Letter, RG&E to NRC Document Control Desk, "Resolution of Generic Letter 87-02, Supplement 1 and Generic Letter 88-20, Supplements 4 and 5 (Seismic Event only),"

dated January 31, 1997 10.

Letter, NRC to RG&E, "Request for Additional Information," dated April6, 1998 Letter, RG&E to NRC Document Control Desk, "Response to NRC Request for Additional Information on the Resolution of Unresolved Safety Issue (USI) A-46, Ginna Nuclear Power Plant," dated May 27, 1998 12.

Letter, NRC to RG&E, "Second Request for Additional Information," December 3, 1998 13.

Letter, RG&E to NRC Document Control Desk, "Response to NRC Second Request for Additional Information on the Resolution of Unresolved Safety Issue (USI) A-46," dated February 2, 1999 14.

Letter, RG&E to NRC Document Control Desk, "Additional Information on Use of GIP Method A, R. E. Ginna Nuclear Power Plant," dated May 25, 1999 "Seismic Review of the Robert E. Ginna Nuclear Power Plant as Part of the Systematic Evaluation Program," R. C. Murray, T. A. Nelson, D. S. Ng, C. Y. Liaw and J. D.

Stevenson, NUREG/CR-1821, December, 1980.

16 GL 87-02, Supplement 1 (SQUG Resolution of USI A-46) and GL 88-20, Supplement 4 (Seismic Events), letter to U.S. Nuclear Regulatory Commission from Robert C. Mecredy of RG&E, dated January 12, 1993 17.

EPRI Report NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1)," dated August 1991 18.

Letter, NRC to Neil Smith (SQUG Chairman), dated June 19, 1998 19.

Letter, NRC to RGKE, "Request for Additional Information," dated March 10, 1999.

20.

Letter, RGLE to NRC, "Response to RAI Regarding Verification of Seismic Adequacy of Mechanical and Electrical Equipment," dated June 7, 1999.