United States Nuclear Regulatory Commission - Protecting People and the Environment

Supplement No. 1 to Generic Letter (GL) 87-02 That Transmits Supplemental Safety Evaluation Report No. 2 (SSER No. 2) on SQUG Generic Implementation Procedure, Revision 2, as Corrected on February 14, 1992 (GIP-2)



TO:         All Unresolved Safety Issue (USI) A-46 Plant Licensees Who Are 
            Members of the Seismic Qualification Utility Group (SQUG) 

SUBJECT:    SUPPLEMENT NO. 1 TO GENERIC LETTER (GL) 87-02 THAT TRANSMITS 
            SUPPLEMENTAL SAFETY EVALUATION REPORT NO. 2 (SSER No. 2) ON SQUG 
            GENERIC IMPLEMENTATION PROCEDURE, REVISION 2, AS CORRECTED ON 
            FEBRUARY 14, 1992 (GIP-2)


Background

GL 87-02, "Verification of Seismic Adequacy of Mechanical and Electrical 
Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46," was 
issued on February 19, 1987.  The generic letter was issued to implement the 
USI A-46  resolution which 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 licensed.  GL 87-02 
requested that all recipients of the letter provide within 60 days of 
receipt of the generic letter a schedule for implementation of the seismic 
verification program at their facilities.  In its April 10 and October 9, 
1987, letters, SQUG, representing its member utilities, committed to a SQUG 
generic program to develop a Generic Implementation Procedure (GIP) for use 
by its members, and requested a deferment of the generic letter's 60-day 
response period until after the NRC issues its final safety evaluation 
report (SER) on the final GIP.  By letter dated November 19, 1987, the staff 
agreed that individual member utility's response to GL 87-02 could be 
deferred until 60 days after the issuance of a final SER on a final GIP.  
(Subsequently, the staff agreed to increase the response period to 120 days 
as described in the corrected Revision 2 of the GIP issued on February 14, 
1992).

Since the issuance of the staff Generic Safety Evaluation Report (GSER) 
dated July 29, 1988, which contains the staff's evaluation of the GIP, 
Revision 0, dated June 1988 for USI A-46, considerable progress has been 
made in the development of technical criteria and procedures, as well as 
some revisions to the licensing issues.  As a result, SQUG incorporated 
parts of the changes and additions into the GIP, Revision 1, dated December 
1988.  The staff reviewed the GIP, Revision 1, and issued Supplemental 
Safety Evaluation Report No. 1 (SSER No. 1), on June 29, 1990.

The SQUG completed the final version of the GIP, Revision 2, as corrected on 
February 14, 1992 (GIP-2), and submitted it to the NRC for review and 
approval



92055190366

                                      2


on February 14, 1992.  The staff has completed its review of the GIP-2.  The 
enclosed SSER No. 2 documents the NRC staff evaluation of GIP-2.  Because 
the GIP criteria and procedures described in Revision 0 and Revision 1 have 
been significantly changed and improved on since the GSER was issued in 
1988, the staff has modified its earlier position on various technical and 
licensing issues.  Therefore, this supplement (SSER No. 2) supersedes all 
previous staff documents such as the GSER and SSER No. 1.

Discussion

This Supplement No. 1. to GL 87-02 transmits the staff SSER No. 2 on the 
SQUG's final GIP, i.e., GIP-2.  As such, SQUG's requested deferment related 
to individual licensee responses to GL 87-02 has ended on the basis that the 
staff has completed the evaluation of the SQUG's final GIP as agreed to with 
the SQUG in the staff's November 19, 1987, letter.  The staff and SQUG 
discussed each of the resolutions in SSER No. 2 needed to close USI A-46.  
Upon completion of the plant-specific walkdown and the third-party audit 
review, the licensee should submit a plant-specific summary report, 
including the results of the third-party audit review and a proposed 
schedule for future modifications and replacements, where appropriate.  Each 
licensee should also provide a completion letter advising the NRC that any 
corrective actions identified in the summary report, or agreed to with the 
staff as a result of other related correspondence, have been completed.  The 
staff will review the licensee's submittals, conduct audits as necessary, 
and issue a simple plant-specific safety evaluation report (SER) addressing 
whether the licensee has complied with its commitments and whether it has 
adequately resolved the USI A-46 issue.  This plant-specific SER will serve 
as the USI A-46 closure document for its respective docket.

Required Response, (10 CFR 50.54f)

Addressees are required to submit, pursuant to the provisions of 10 CFR 
50.54(f), the following information within 120 days of the date of this 
Supplement No. 1 to GL 87-02:

1. A statement whether you commit to use both the SQUG commitments and the 
   implementation guidance provided in GIP-2 as supplemented by the SSER No. 
   2 for the resolution of USI A-46.  In this case, any deviation from 
   GIP-2, as supplemented by the SSER No. 2, must be identified, justified, 
   and documented.  If you do not make such a commitment, you must provide 
   your alternative for responding to GL 87-02.

2. A plant-specific schedule for the implementation of the GIP and 
   submission of a report to the staff that summarizes the results of the 
   USI A-46
                                      3



   review, if you are committing to implement GIP-2.  This schedule shall be 
   such that each affected plant will complete its implementation and submit 
   the summary report within 3 years after the issuance of the SSER No. 2, 
   unless otherwise justified.

3. The detailed information as to what procedures and criteria were used to 
   generate the in-structure response spectra to be used for USI A-46 as 
   requested in the SSER No. 2.  The licensee's in-structure response 
   spectra are considered acceptable for USI A-46 unless the staff indicates 
   otherwise during a 60-day review period.

Your response must be submitted under oath or affirmation and must be 
addressed to the U.S. Nuclear Regulatory Commission, ATTN:  Document Control 
Desk, Washington, D.C. 20555.  In addition, a copy of your response must be 
submitted to the appropriate regional administrator.  The justification for 
this information request under 10 CFR 50.54(f) continues to be the same as 
that for GL 87-02.

Backfit Discussion

The staff's request for a plant-specific schedule and information related to 
the implementation of the resolution of USI A-46 was considered to be a 
backfit under 10 CFR 50.109.  A backfit analysis was completed and described 
in NUREG-1211, "Regulatory Analysis for Resolution of Unresolved Safety 
Issue A-46, Seismic Qualification of Equipment in Operating Plant," dated 
February 19, 1987.  Although the estimated cost to a licensee excluding the 
cost of repairs to anchorages and supports of equipment in implementing the 
USI A-46 program has since increased approximately 25% in total cost, the 
benefits of implementing the program have also greatly increased as 
documented in the package of information prepared for the Committee to 
Review Generic Requirements, which reviewed this Supplement No. 1 to GL 
87-02 at the Meeting No. 212.  The staff has placed the package of 
information in the Public Document Room for public availability.

This request is covered by Office of Management and Budget Clearance Number 
3150-0011, which expires May 31, 1994.  The estimated average number of 
burden hours is 120 person hours per owner response, including the time 
required to assess the new recommendations, search data sources, gather and 
analyze the data, and prepare the required letters.  These estimated average 
burden hours pertain only to the identified response-related matters and do 
not include the time for actual implementation of the requested action.  
Comments on the accuracy of this estimate and suggestions to reduce the 
burden may be directed to Ronald Minsk, Office of the Information and 
Regulatory Affairs (3150-0011), NEOB-3019, Office of Management and Budget, 
Washington, D.C.  20503, and to the U.S. Nuclear Regulatory Commission, 
Information and Records Management Branch, (MNBB 7714), Division of 
Information Support Service, Office of Information and Resources Management, 
Washington, D.C. 20555. 

                                      4



If you have any questions about this matter, please contact the NRC 
technical contact or the lead project manager listed below.

                        Sincerely,

   
                        James G. Partlow
                        Associate Director for Projects
                        Office of Nuclear Reactor Regulation

Enclosures:  
1.  SSER No. 2
2.  Listing of Recently Issued Generic Letters

cc w/enclosures:
Seismic Qualification Utility Group
ATTN:  Richard Schaffstall
Electric Power Research Institute
1019 19th Street, NW
Washington, DC 20336


Technical Contact: Pei-Ying Chen (301) 504-2789

Lead Project Manager: Patrick Sears (301) 504-2021







                 SUPPLEMENTAL SAFETY EVALUATION REPORT NO. 2


                                     ON


                    SEISMIC QUALIFICATION UTILITY GROUP'S

                GENERIC IMPLEMENTATION PROCEDURE, REVISION 2,

                         CORRECTED FEBRUARY 14, 1992



                                     FOR



                   IMPLEMENTATION OF GL 87-02 (USI A-46),


                VERIFICATION OF SEISMIC ADEQUACY OF EQUIPMENT

                      IN OLDER OPERATING NUCLEAR PLANTS

.

                              TABLE OF CONTENTS
                                                                     Page
BACKGROUND                                                              1
GENERAL DISCUSSION                                                      3
GENERAL EVALUATION . . .                                                3
DETAILED DISCUSSION AND EVALUATION                                      5

I     Licensing and Implementation Guidelines                           5
      I.1.0       Introduction                                          5
      I.1.1       Background                                            5
      I.1.2       Purpose of the GIP                                    6
      I.1.3       GIP Commitments and Guidance                          6
      I.2.0       Issues and Positions                                  6
      I.2.1       Introduction                                          6
      I.2.2       Interpretation and Guidelines                         7
      I.2.3       Compliance With Guidelines                            7
      I.3.0       Revisions to the GIP                                  9

II    Generic Procedure for Plant-Specific Implementation               9
      II.1        Introduction                                          9
      II.2        Seismic Evaluation Personnel                          9
      II.3        Identification of Safe-Shutdown Equipment            10
      II.4        Screening Verification and Walkdown                  12
                
            II.4.0  Introduction                                       12
            II.4.1      SQUG Commitments                               13
            II.4.2      Seismic Capacity Compared to Seismic Demand    13
            II.4.3      Equipment Class Similarity and Caveats         16
            II.4.4      Anchorage Adequacy                             16
      
      II.5        Outlier Identification and Resolution                23
      II.6        Relay Functionality Review                           24
      II.7        Tanks and Heat Exchangers Review                     26
      II.8        Cable and Cable Raceway Review                       29
      II.9        Documentation                                        31
      II.10       Reference                                            31

III   Appendices. . . .                                                32
      III.1       Appendix A, Procedure for Identification of          32
                              Safe Shutdown Equipment
      III.2       Appendix B, Summary of Equipment Class Descriptions  32
                              and Caveats
      III.3       Appendix C, Anchorage Data                           33
      III.4       Appendix D, Seismic Interaction                      33
      III.5       Appendix E, Preparatory Work Prior to Walkdown       33
      III.6       Appendix F, Screening Walkdown Plan                  34
      III.7       Appendix G, Screening Evaluation Work Sheets         34

CONCLUSION. . . . . . .                                                34
REFERENCES. . .   . . . .                                              37



.

                 SUPPLEMENTAL SAFETY EVALUATION REPORT NO. 2
                  ON SEISMIC QUALIFICATION UTILITY GROUP'S
                GENERIC IMPLEMENTATION PROCEDURE, REVISION 2,
                        CORRECTED FEBRUARY 14, 1992, 
              FOR IMPLEMENTATION OF GL 87-02, USI A-46 PROGRAM

              VERIFICATION OF SEISMIC ADEQUACY OF EQUIPMENT IN
                       OLDER OPERATING NUCLEAR PLANTS

BACKGROUND

In December 1980, the Nuclear Regulatory Commission (NRC) designated 
"Seismic Qualification of Equipment in Operating Plants" as an unresolved 
safety issue (USI).  The safety issue of concern was that equipment in 
nuclear plants for which construction permit (CP) applications had been 
docketed before about 1972 had not been reviewed according to the 
then-current (1980-81) licensing criteria for seismic qualification of 
equipment (i.e., Regulatory Guide (RG) 1.100; Institute of Electrical and 
Electronics Engineers (IEEE) Standard 344-1975, and Standard Review Plan 
(SRP) Section 3.10 (NUREG-0800, July 1981)).  Therefore, the seismic 
adequacy of the equipment in these older plants may be questionable 
regarding their ability to survive and function in the event of a 
safe-shutdown earthquake (SSE).  Equipment in plants with a CP application 
docketed after about 1972 were qualified according to the then-current 
licensing criteria and licensee compliance has been audited by the NRC 
staff.  All operating plants for which equipment seismic qualification could 
not be verified to meet the intent of then-current licensing criteria are 
subject to the implementation provisions outlined in Generic Letter (GL) 
87-02, "Verification of Seismic Adequacy of Mechanical and Electrical 
Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46" 
(Reference 1).  These plants are identified as "USI A-46 plants" and are 
listed in Table A in Section II.4.2 of this report. 

The applicable portions of the NRC's regulations governing the seismic 
design of nuclear power plants require that structures, systems, and 
components important to safety be designed to withstand the effects of 
earthquakes, and that those systems and components be designed to perform 
their intended safety functions (Appendix A to 10 CFR Part 50).  Appendix A 
to 10 CFR Part 100, which was published in the Federal Register on November 
13, 1973 (38 FR 31281) and became effective December 13, 1973, requires that 
the engineering method, used to insure that the required safety functions of 
such structures, systems, and components are maintained during and after an 
SSE, must involve the use of either a suitable dynamic analysis or a 
suitable qualification test.  This engineering method is termed "seismic 
qualification method" for the purpose of licensing requirements.  No 
explicit provisions within the regulations permit the use of experience data 
as a means for seismic qualification.  However, the NRC has determined that 
requiring those older operating plants to comply with the then-current 
licensing requirements was not practicable because a literal application of 
those criteria to older operating plants could require extensive 
modifications of those facilities that could not be justified from the 
cost-benefit standpoint. 

Although no explicit provisions within the regulations permit the use of 
experience data as a means for seismic qualification, the NRC concluded that 
the use of earthquake experience data, with appropriate restrictions and 
.

caveats, supplemented by some test results to verify the seismic adequacy of 
equipment within certain specified earthquake motion bounds, represented the 
most reasonable and cost-effective means of ensuring that the purpose of the 
NRC regulations related to seismic design can be satisfied for those plants.  
Therefore, for USI A-46 plants only, rather than requiring compliance with 
then-current criteria for seismic qualification of equipment, the staff 
requested  the USI A-46 licensees only to verify the seismic adequacy of 
equipment in these plants as reflected in the title and the intent of GL 
87-02 (Reference 1).  One of the programmatic restrictions excludes using 
earthquake experience data to verify the seismic adequacy of structures and 
piping.

To address the USI A-46 issue, some of the affected utilities formed the 
Seismic Qualification Utility Group (SQUG) in 1982.  In 1983, the SQUG 
proposed the formation of a panel of consultants, the Senior Seismic Review 
and Advisory Panel (SSRAP), to independently assess and review the viability 
of using earthquake experience data and test data to demonstrate equipment 
ruggedness, and to provide expert advice and consultation.  The SQUG 
subsequently developed the "Generic Implementation Procedure (GIP) for 
Seismic Verification of Nuclear Plant Equipment" for its members to use.  
The SQUG submitted the GIP, Revision 0 (GIP-0), dated June 1988 (Reference 
2) and the related documents and reports supporting GIP-0 to the NRC staff 
for review.  The staff reviewed these documents and issued a Generic Safety 
Evaluation Report (GSER) on July 29, 1988 (Reference 3), recognizing that 
not all sections of GIP-0 had been developed at that time.

In contrast to the IEEE Standard 344 qualification approach, which in the 
past has relied on analysis or testing of each item of equipment, the GIP 
methodology relies primarily on the use of existing earthquake and test 
experience data to verify the seismic adequacy of generic classes of 
equipment.  By convention, the IEEE Standard 344 procedures have been termed 
"equipment seismic qualification," while the USI A-46 procedures have been 
termed "equipment seismic adequacy verification." 

In December 1988, Revision 1 of the GIP (GIP-1) (Reference 4) was submitted 
for NRC staff review.  GIP-1 contained essentially the same technical topics 
as GIP-0 except that a new section was  added for evaluating tanks and heat 
exchangers and some information was added for resolving outstanding issues.  
While the staff was reviewing GIP-1, the SQUG was making significant changes 
to Part II which meant that Part II of GIP-1 would be virtually obsolete.  
Therefore, the staff focused its evaluation of GIP-1 primarily on Part I.  
This evaluation can be found in the Supplemental Safety Evaluation Report 
(SSER) No. 1 (Reference 5). 

In September 1990, the SQUG submitted Revision 2 of the GIP (Reference 6).  
The staff reviewed Revision 2 and commented on it in March 1991 (Reference 
7).  In response to these comments and to subsequent discussions with the 
NRC staff, the SQUG further revised Revision 2, and resubmitted it with 
corrected pages in February 1992 (Reference 8).  This supplement (SSER No. 
2) presents the results of the NRC staff's evaluation of this latest 
revision.  For the remainder of this report, "GIP-2" refers to Reference 8.
.

Because the criteria and procedures described in GIP-0 and GIP-1 have been 
significantly changed and improved since the GSER was issued in 1988, the 
staff has modified its earlier positions on various technical and licensing 
issues.  Therefore, this supplement (SSER No. 2) supersedes all such 
previous staff documents, i.e, the GSER and SSER No. 1.

This supplement begins with a general discussion and evaluation of the 
overall GIP-2 document, followed by a detailed discussion and evaluation of 
specific sections of GIP-2.  Where it is applicable, the staff discusses 
clarifications, interpretations, positions and exceptions.  The 
clarifications, interpretations, positions and exceptions are not 
specifically labeled in the main text of this report, but all exceptions are 
specifically identified in the final conclusion section.

GENERAL DISCUSSION

GIP-2 is divided into two major parts:  Part I discusses the related 
licensing and implementation issues for the USI A-46 program; Part II 
contains the technical information necessary for the implementation of the 
program.  Part II has 10 sections.  They are:

1.    Introduction 
2.    Seismic Evaluation Personnel 
3.    Identification of Safe Shutdown Equipment 
4.    Screening Verification and Walkdown 
5.    Outlier Identification and Resolution 
6.    Relay Functionality Review 
7.    Tanks and Heat Exchangers Review 
8.    Cable and Conduit Raceway Review 
9.    Documentation 
10.   References

GIP-2 provides the general guidelines in these 10 sections with detailed 
procedures, technical data, and implementation work sheets given in seven 
appendices.  These are:

Appendix A  Procedure for Identification of Safe Shutdown Equipment 
Appendix B  Summary of Equipment Class Descriptions and Caveats 
Appendix C  Anchorage Data 
Appendix D  Seismic Interaction 
Appendix E  Preparatory Work Prior to Walkdown 
Appendix F  Screening Walkdown Plan 
Appendix G  Screening Evaluation Work sheets 

Parts, sections, and appendices of GIP-2 are discussed in more detail in 
this SSER No. 2 in the section "Detailed Discussion and Evaluation." 

GENERAL EVALUATION

In general, the NRC staff finds GIP-2 to be a very useful working document 
for implementing the USI A-46 program.  The information contained in GIP-2 
is 
.

generally acceptable to the staff for a plant-specific implementation of USI 
A-46.

The staff discusses clarifications, interpretations, exceptions, and 
positions in the section "Detailed Discussion and Evaluation," by 
referencing specific parts, sections, or appendices of GIP-2.  The staff 
clarifications and exceptions that are general in nature and that apply to 
the entire GIP-2 are listed as follows:

1. The staff considers GIP-2 acceptable (with the clarifications, 
   interpretations, exceptions, and positions identified in this SSER No. 2) 
   for verifying seismic adequacy of equipment in USI A-46 plants only.  The 
   NRC staff expects that, when weaknesses in existing equipment are 
   identified as a result of the implementation of USI A-46, licensees will 
   take appropriate corrective actions, including modifications or upgrades, 
   if necessary, to ensure that those equipment possess an adequate level of 
   seismic safety.

2. The staff considers GIP-2 to be a method for verifying the seismic 
   adequacy of equipment rather than a seismic qualification method.  This 
   is because the criteria and many of the practices proposed in GIP-2 are 
   not equivalent to current seismic qualification requirements.  Examples 
   include the following:  (1) A vast amount of the technical information 
   that is given in GIP-2 was gathered in a very general way rather than for 
   each item of equipment, and was based on many subjective judgments and 
   opinions; (2) The implementation methodology proposed in GIP-2 allows the 
   review engineers to resolve major issues on the basis of their judgments, 
   in some cases without requiring the engineers to justify or document the 
   basis for these judgments, rather than on the basis of such standard 
   engineering practices as calculations and testing; (3) The damping values 
   used in GIP-2 are, in general, higher than those provided in the current 
   version of R.G. 1.61; (4) The practice proposed in GIP-2 for evaluating 
   safe-shutdown paths and identifying safe-shutdown equipment differs from 
   current requirements in that safety grade equipment is not required to be 
   available and it is not necessary to cool the reactor beyond hot shutdown 
   conditions, whereas current designs require safety grade systems to cool 
   the reactor from normal operating conditions to cold shutdown; and (5) 
   The practice of allowing the "rule of the box" (GIP-2, page 3-16) and 
   spot checking of the device mounting in a cabinet also differs from 
   current practice, which requires such activities as testing, inspection, 
   documentation, and corrective actions to be covered by a 10 CFR Part 50, 
   Appendix B, quality assurance program.

   On the basis of the differences between current seismic qualification 
   requirements and the criteria and the procedures provided in GIP-2, the 
   NRC staff does not consider the USI A-46 methodology given in GIP-2 to be 
   a "seismic qualification" procedure.  Rather, the staff considers the 
   GIP-2 methodology to be a seismic adequacy verification procedure which 
   was developed based on generic equipment earthquake experience data 
   supplemented by generic equipment test data.  The implementation of the 
   GIP-2 approach for USI A-46 plants provides safety enhancement, in 
   .   

   certain aspects, beyond the original licensing bases.  Therefore, GIP-2 
   methodology is an acceptable evaluation method, for USI A-46 plants only, 
   to verify the seismic adequacy of the safe-shutdown equipment and to 
   satisfy the pertinent equipment seismic requirements of General Design 
   Criterion 2 and the purpose of the NRC regulations relevant to equipment 
   seismic adequacy including 10 CFR Part 100.

3. The term "licensee," as used in GIP-2, refers only to the licensee of a 
   plant in the USI A-46 program.

4. Some statements were made in Reference 5 of Part II of GIP-2 about the 
   aging effects of the database equipment.  The NRC staff considers that 
   the scope of the USI A-46 program excludes the issue of environmental 
   qualification of equipment in operating plants, because this issue was 
   addressed by the implementation program under 10 CFR 50.49.  GIP-2 does 
   not address the aging effects of equipment by systematic collection of 
   quantitative data on the earthquake experience; therefore, the staff will 
   not accept any claim that the experience data collected by the SQUG for 
   the USI A-46 program adequately addressed the aging effects of equipment, 
   as one might incorrectly interpret from the related statement on page 13, 
   Reference 5 of Part II of GIP-2. 

DETAILED DISCUSSION AND EVALUATION

In the details of the staff's evaluation of GIP-2, which follows, parts, 
sections, or appendices are briefly discussed, and the staff's positions, 
clarifications, interpretations, and exceptions are presented.

I  LICENSING AND IMPLEMENTATION GUIDELINES

Part I of GIP-2 describes the genesis of the USI A-46 program and discusses 
the role of the GIP in resolving the unresolved safety issue.  This part 
considers several issues and describes SQUG positions on several aspects 
related to licensing and implementation guidelines.  These aspects include, 
among other things, the interpretation of GIP-2 guidelines, the compliance 
with regulations, the selection of equipment, and any future revisions of 
GIP-2.  The staff finds Part I acceptable subject to the following: 

I.1.0  Introduction

I.1.1  Background

The second paragraph of Section 1.1 of Part I of GIP-2 states that "the 
purpose of USI A-46 is to verify this conclusion" which is "that there is 
adequate seismic capacity of properly anchored equipment in older operating 
plants."   Although this conclusion is expected to be correct in general, 
there may be some pieces of equipment for which proper anchorage alone does 
not demonstrate seismic adequacy.  For all equipment within the USI A-46 
scope, the licensee is responsible for verifying all aspects of seismic 
adequacy of the equipment in accordance with GIP-2.
.

I.1.2  Purpose of the GIP

1. The third paragraph of Section 1.2 of Part I states that "Because the NRC 
   will document its evaluation of the GIP in a safety evaluation report 
   (SER), the GIP provides an NRC-accepted method to verify the seismic 
   adequacy of equipment...."  The staff concurs with this statement, 
   provided that GIP-2 is used in its entirety in conjunction with and 
   supplemented by the clarifications, interpretations, and exceptions 
   identified in this supplement, and that the application of the GIP is 
   limited to USI A-46 plants only.

2. Section 1.2 of Part I states, "Every aspect of the Generic Letter 
   Procedure has been fully considered in development of the GIP.  
   Therefore, licensees will be guided by the GIP.  By satisfying the 
   provisions of the GIP, licensees will have fully satisfied the guidance 
   of the Generic Letter...."  This is generally acceptable to the staff for 
   GIP-2.  However, deviations from GIP-2 and the SSER No. 2 without the 
   staff's prior approval may result in the licensee not fully satisfying 
   the provisions of GL 87-02.

I.1.3  GIP Commitments and Guidance

1. The second paragraph of Section 1.3 of Part I states that "USI A-46 
   licensees may use the GIP guidance or may substitute clearly equivalent 
   methods without prior notification of the NRC...."  The staff's position 
   is that if licensees use other methods that deviate from the criteria and 
   procedures as described in SQUG commitments and in the implementation 
   guidance of GIP-2 without prior NRC staff approval, the method may not be 
   acceptable to the staff and, therefore, may result in a deviation from 
   the provisions of GL 87-02 as stated in item 2 in Section I.1.2 above.

2. The third paragraph of Section 1.3 of Part I states that "submittals 
   which commit to the entire GIP...shall be regarded as accepted by the 
   Staff upon docketing...."  The staff concurs with this statement provided 
   that the licensee commits to the entire GIP-2 as supplemented by the 
   clarifications, interpretations, and exceptions identified in this 
   supplement.

I.2.0  Issues and Positions

I.2.1  Introduction

The third paragraph of Section 2.1 of Part I refers to SQUG documents (e.g. 
References 9 and 12 of Part I of GIP-2) that summarize the resolution 
histories of many issues.  The staff recognizes the importance of these 
documents.  Although all of these many issues are resolved with or without 
conditions or clarifications, these documents reflect SQUG's perception of 
the resolution.  The final resolutions of all issues are contained in the 
GIP-2 as supplemented by this SSER No. 2.

.

I.2.2  Interpretation and Guidelines

For a meaningful third-party audit (Section 2.2.7 of Part I), the NRC 
expects that the auditor(s) should have broad engineering experience and 
have completed the SQUG developed training course on seismic adequacy 
verification of equipment in operating nuclear power plants.  This is 
because the third-party audit will involve substantially less time and 
effort than the original walkdown and analyses.  Thus, the auditor(s) should 
have sufficient qualification and experience to be able to assess the 
adequacy of the entire plant-specific implementation program during the 
limited time of the audit.

Additionally, to provide a desired degree of assurance concerning the 
effectiveness of the third-party review, a process for inter-plant 
information exchange and coordination should be implemented to collect, 
evaluate, and disseminate generic problems, questions, and lessons learned 
during the USI A-46 plant-specific walkdowns and third-party reviews to all 
member utilities in a timely manner.  The responsibility of carrying out the 
above-mentioned process may be charged to the cognizant industry 
organization as stated in Section I.3.0 of this supplement.

I.2.3  Compliance With Regulations

1. Section 2.3.3 of Part I, Revision of Plant Licensing Bases, states that, 
   "a USI A-46 licensee, in accordance with 10 CFR . 50.59, may revise the 
   plant licensing bases to reflect that the USI A-46 (GIP) methodology may 
   henceforth be used as the methodology for verifying the seismic adequacy 
   of mechanical and electrical equipment within the scope of equipment 
   covered by the GIP...."  The staff recognizes that a licensee may revise 
   its licensing basis in accordance with 10 CFR 50.59 to reflect the 
   acceptability of the USI A-46 (GIP) methodology for verifying the seismic 
   adequacy of electrical and mechanical equipment covered by the GIP.  The 
   staff's approval of the implementation of the GIP does not relieve the 
   licensees from the requirement to address all aspects of unreviewed 
   safety questions as specified in 10 CFR 50.59 (for example, those plants 
   where the FSAR has specified damping values which differ from the GIP.)  
   The staff understands the word "henceforth" to mean, based on SQUG GIP-0 
   (page 5 of Part I), "after issuance of a final, plant-specific SER 
   resolving USI-A-46."  If this is not the case, the staff requests 
   licensees intending to change their licensing bases prior to receipt of 
   the plant-specific SER to inform the staff in their 120-day response 
   letters.

2. In Section 2.3.3 of Part I, Example 2 and Example 4 may imply that the 
   seismic requirements (RG 1.100, Revision 1) for RG 1.97 instrumentation 
   may be changed to the GIP seismic methodology under 10 CFR 50.59.  The 
   staff has stated, and the SQUG has previously acknowledged, that any 
   previous commitments, such as for RG 1.97 and TMI Action Plan Item 
   II.F.2, are not superseded by the resolution methods of the GIP.  For 
   Category 1 equipment, as described in RG 1.97, the staff agrees that the 
   seismic qualification requirements (RG 1.100, Revision 1) will resolve 
   the USI A-46 requirements for that equipment.  The Category 2 and 
   Category 3 equipment as described in RG 1.97 have no specific seismic 
   .   

   qualification provisions.  Therefore, if that equipment is used as part 
   of the USI A-46 safe-shutdown equipment, it will need to be verified for 
   seismic adequacy using GIP-2 methods or by an acceptable seismic 
   qualification method.

3. Section 2.3.3 of Part I is acceptable to the staff subject to the 
   addition of the following phrase to the last sentence of Example 5:  "... 
   for matters related to verifying the seismic adequacy of electrical and 
   mechanical equipment." 

4. Section 2.3.4 of Part I describes the criteria and procedures for future 
   modification and for new and replacement equipment.  The staff position 
   is that these criteria and procedures may be applied to new and 
   replacement equipment on a case-by-case (i.e., plant-specific and 
   equipment-specific) basis only and with the provisions that the seismic 
   evaluations are performed in a systematic and controlled manner so as to 
   ensure that new or replacement items of equipment are properly 
   represented in the earthquake experience or generic testing equipment 
   classes, and that applicable caveats are met.  In particular, each new or 
   replacement item of equipment and parts must be evaluated for any design 
   changes that could reduce its seismic capacity from that reflected by the 
   earthquake experience or generic testing equipment classes, and these 
   evaluations must be documented.  These criteria and procedures as 
   described are acceptable for verifying the seismic adequacy of 
   commercial-grade equipment to be dedicated for safety-related purposes; 
   but, for other (non-seismic) critical characteristics of equipment to be 
   dedicated, licensees are referred to such applicable guidance and 
   requirements as GL 89-02, GL 89-09, and GL 91-05, which include 
   applicable criteria of 10 CFR Part 50, Appendix B.

   The staff normally would require that new or replacement equipment be 
   qualified in accordance with plant-specific licensing commitments or 
   current criteria (e.g., 10 CFR 50.49) unless the licensee can justify the 
   use of other acceptable qualification methods.  As a result of the 
   backfit analysis for the USI A-46 program, the staff determined that the 
   use of USI A-46 approach provides  adequate level of safety and that it 
   was not cost-justifiable for the safety benefit gained to demonstrate the 
   seismic qualification of equipment in these older operating plants by 
   using rigorous current qualification requirements.  Therefore, the 
   resolution as described in GL 87-02 and NUREG-1211, "Regulatory Analysis 
   for Resolution of Unresolved Safety Issue A-46, 'Seismic Qualification of 
   Equipment in Operating Plants'," was that the criteria and procedures 
   described herein are determined to be an acceptable evaluation method for 
   verifying the seismic adequacy of the equipment in USI A-46 plants 
   including future modifications and replacement equipment in these plants. 

   The backfit analysis described in NUREG-1211 did not specifically address 
   new equipment.  However, the staff agrees that it is impractical and 
   inconsistent with the USI A-46 philosophy to require that new equipment 
   shall meet current seismic qualification requirements, whereas the 
   seismic adequacy of all other safe shutdown equipment (which will 
   .   

   presumably encompass the large majority of all safe shutdown equipment in 
   the plant) is verified through the USI A-46 procedures.  Therefore, the 
   criteria and procedures described herein are determined to be an 
   acceptable evaluation method for verifying the seismic adequacy of new 
   equipment in USI A-46 plants.


I.3.0  Revisions to the GIP

Section 3.0 of Part I mentions that the earthquake experience or generic 
testing equipment classes will be periodically modified by a cognizant 
industry organization as new information becomes available.  Although the 
staff does not intend to review every detail of the information collected, 
the suggested cognizant industry organization should submit, for NRC staff 
review and approval, a procedure for evaluating the acceptability of new 
data and a procedure for updating and revising GIP-2 and subsequent 
revisions, based on new information including the lessons learned during the 
USI A-46 plant walkdowns.

II  GENERIC PROCEDURE FOR PLANT-SPECIFIC IMPLEMENTATION

Part II of GIP-2 which provides the implementation guidelines for the USI 
A-46 program, contains 10 sections and 7 appendices.  These sections and 
appendices are given below.

II.1  Introduction

Section 1 of Part II describes the purpose, background and approach used in 
GIP-2.  This section also introduces other sections and discusses to some 
extent the following subjects:

   seismic evaluation personnel 
   identification of safe shutdown equipment 
   screening verification and walkdown
   outlier identification and resolution 
   relay functionality review 
   tanks and heat exchangers review 
   cable and conduit raceway review 
   documentation


II.2  Seismic Evaluation Personnel

Discussion

Section 2 of Part II defines the responsibilities and qualifications of the 
engineers who will perform seismic evaluations of the equipment.  The 
systems engineers will develop the list of equipment required for safe 
shutdown.  The systems engineer should be a degree engineer, or equivalent, 
and should have had extensive experience with, and broad understanding of, 
the systems, equipment, and procedures of the plant.  The seismic capability 
engineers will conduct the walkdowns and assess the seismic adequacy of 
safe-shutdown 
.

equipment.  The seismic capability engineers should be degree engineers, or 
equivalent, who have completed a SQUG-developed training course on seismic 
adequacy verification of nuclear power plant equipment.  These engineers 
should have at least 5 years of experience in earthquake engineering 
applicable to nuclear power plants and in structural or mechanical 
engineering.  At least one of the seismic capability engineers on each of 
the seismic review teams should be a licensed professional engineer to 
ensure that there is a measure of accountability and personal responsibility 
in making the equipment seismic adequacy determination.  The relay reviewers 
will perform the functionality review of the relays with the safe-shutdown 
functions.  The lead relay reviewer should be a degree, or equivalent, 
electrical engineer with some electrical engineering experience who is 
familiar with the relay functionality review procedure described in Section 
6 of Part II and Reference 8 of GIP-2.  The lead relay reviewer should 
successfully complete the SQUG-developed relay training course.  The plant 
operations staff will review the safe-shutdown equipment list and assist the 
seismic capability engineers and the relay review team.  The plant 
operations personnel should have experience in the specific plant being 
seismically verified.

Evaluation and Conclusion

Based on the above discussions, the staff finds the criteria for qualifying 
those individuals responsible for implementing the GIP-2 procedure 
acceptable in that the required qualifications are adequate to assure that 
the GIP-2 is performed in an acceptable fashion.  

The staff acknowledges that these responsible individuals must exercise 
judgment to implement the USI A-46 program.  The review engineers should 
utilize the technical information in the GIP-2 and the reference documents 
to the maximum extent practicable in determining the seismic adequacy of 
equipment.  Where judgements are needed to make these determinations, the 
assumptions and basis for the judgmental conclusions should be documented as 
required in GIP-2 or identified in this supplement.

II.3  Identification of Safe-Shutdown Equipment

Discussion

Section 3 of Part II describes the overall method for identifying the 
equipment needed to achieve and maintain safe-shutdown conditions in a 
nuclear plant.  The SQUG commitments, general criteria and governing 
assumptions, scope of equipment, safe-shutdown functions, safe-shutdown 
alternatives, identification of equipment, operations department review, and 
documentation are the major subjects discussed in this section.  Loss of 
offsite power as a result of SSE is assumed, and the systems selected for 
the safe shutdown should not be dependent on a single piece of equipment 
whose failure would preclude a safe shutdown.  Based on these assumptions 
and others as specified in GIP-2, the licensee will use the following 
two-stage approach to identify the equipment needed to achieve and maintain 
a safe-shutdown condition:

.

1. The licensee will select a safe-shutdown path which would ensure that the 
   four essential safe-shutdown functions listed below can be accomplished 
   following an SSE.  The functions are:

      reactor reactivity control
      reactor coolant pressure control
      reactor coolant inventory control
      decay heat removal

2. After identifying the safe-shutdown path, the licensee will identify the 
   individual items of equipment required to accomplish the four essential 
   safe-shutdown functions.

Evaluation and Conclusion

The staff finds the proposed two-stage approach adequate for identifying the 
equipment needed to achieve and maintain a safe-shutdown condition.  
Therefore, the staff concludes that Section 3 of Part II and Appendix A of 
GIP-2 are acceptable subject to the following:

1. Regarding the safe-shutdown equipment list (SSEL), the "safe shutdown" is 
   defined as bringing the plant to, and maintaining it in, a hot shutdown 
   condition during the first 72 hours following an SSE (i.e. within 
   72 hours, the plant is cooled down to the "hot shutdown" condition in 
   accordance with the plant-specific Technical Specifications).  The intent 
   of this position is to have pressurized-water reactors (PWRs) lower their 
   temperature and pressure within 72 hours to the point at which residual 
   heat removal (RHR) equipment could be used, but does not necessarily 
   require RHR equipment to be included on the SSEL.  The staff does not 
   intend to require plants to cool down faster than their original design 
   capability or technical specification limits.  Therefore, if a licensee 
   cannot achieve hot shutdown at a plant within 72 hours, the licensee 
   should discuss with, and obtain prior written consent from the staff on a 
   case-by-case basis before implementing the USI A-46 program.

2. All facilities have Emergency Operating Procedures (EOPs) which address 
   actions in the event of an accident.  As stated in Section 3.2.8 of GIP-
   2, the staff expects that plant operators should be trained in the use of 
   existing normal shutdown procedures or symptom-based EOPs which would be 
   used if a safe shutdown earthquake were to occur.  The compatibility of 
   these procedures with the USI A-46 safe shutdown equipment list should be 
   verified by the plant Operations Department, and the results included in 
   the operator training program.  This will ensure that the shutdown path 
   selected for USI A-46 (and equipment included in the SSEL) is a 
   legitimate safe-shutdown path consistent with plant procedures and 
   operator training.  

3. Because some components, such as those made of cast iron, are brittle and 
   are more vulnerable to earthquake damage, any such equipment identified 
   during the walkdown shall be specifically evaluated for seismic adequacy.  
.

4. With regard to Section 3.3.2 of Part II "Exclusion of NSSS Equipment," 
   the staff finds that the technical basis provided in Reference 17 of 
   GIP-2 is acceptable for excluding those items of equipment listed in 
   Section 3.3.2 with the exception of safety-relief valves.  The NSSS 
   equipment exclusion given in Section 3.3.2 of GIP-2 does not apply to 
   safety-relief valves included in the USI A-46 scope because Reference 17 
   of GIP-2 does not provide a basis for excluding the safety-relief valves 
   from the USI A-46 scope.

5. Section 3.3.3 of Part II requires that any equipment needed for safe 
   shutdown be evaluated for relay chatter.  For example, even if equipment 
   such as a pump is itself seismically rugged, the effects of relay chatter 
   on the electric power and instrumentation and control circuits still need 
   to be evaluated to ensure the equipment functionality.

II.4   Screening Verification and Walkdown

Discussion

Section 4 of Part II describes the screening verification and walkdown 
procedures that will be implemented to verify the seismic adequacy of the 
equipment.  In summary, the licensee should (1) compare the seismic capacity 
with the demand, (2) satisfy the caveats of the respective databases, (3) 
check the anchorages for adequacy, and (4) consider the seismic 
interactions.

Evaluation

Section 4 of Part II provides the first level of screening of the equipment 
required for safe shutdown for its seismic adequacy.  GIP-2 also provides 
criteria and screening procedures for five types of anchorages which have 
been used extensively in the nuclear power plants to secure equipment.  The 
criteria provide guidance for determining the seismic load acting on, and 
the allowable load of, individual anchors to be calculated and compared.  
Anchors will be classified as outliers if the loads acting on the anchors 
exceed their allowable parameters.  Some anchors could be identified as 
outliers during visual inspection of the screening procedures.  The 
evaluation of screening verification and walkdown follows in Sections 
II.4.2, II.4.3, and II.4.4 of this supplement.

Conclusion

The staff has reviewed the screening procedures and criteria.  Based on the 
evaluations and findings described in Sections II.4.2, II.4.3, and II.4.4 
below, the staff concludes that the screening procedures and criteria are 
adequate and acceptable only for verifying seismic adequacy of equipment in 
USI A-46 plants, subject to the staff clarifications, interpretations, 
exceptions and positions described in the sections that follow.

II.4.0  Introduction

This section provides a summary and organization of Section 4 of Part II of 
GIP-2.  The staff has no comment on this section.
.

II.4.1  SQUG Commitments

Section 4.1 of Part II provides SQUG general commitments in the areas of 
screening verification and walkdown procedures.  The SQUG commitments and 
the implementation guidance of GIP-2 were developed to form an integral part 
to satisfy the guidance of GL 87-02.  Therefore, the staff position is that 
the licensee must commit to both the SQUG commitments and the use of entire 
implementation guidance provided in GIP-2, unless otherwise justified to the 
staff as described in GIP-2 and this supplement (see Item 2 in Section I.1.2 
and Item 1 in Section I.1.3 of this supplement).

II.4.2  Seismic Capacity Compared to Seismic Demand

1. Section 4.2 of Part II maintains that "... 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..." (page 4-10 of GIP-2).  The NRC staff 
   cautions that because an equipment assembly (e.g., electrical cabinet 
   lineup) may consist of many subassemblies, each manifesting its 
   fundamental mode of vibration at different frequencies, the GIP-2 
   approach may be non-conservative unless all such frequencies are 
   determined with high confidence.  In addition, unless the equipment is 
   tested with a high-level vibratory input, the fundamental frequency is 
   extremely difficult to estimate, especially for complex structured 
   equipment.  Therefore, the staff position is that the capacity spectrum 
   should envelope the demand spectrum over the entire frequency range 
   unless the lowest natural frequency of the equipment, the door panels, or 
   internal structures and components as described in Section 4.2 of GIP-2 
   can be conservatively established (see Item 3 of Section III.7 of this 
   supplement).

2. In regards to the comparison of seismic capacity to demand in methods 
   A.1, A.2, (page 4-14 of GIP-2), and B.1, B.2 (page 4-17 of GIP-2), the 
   SQUG proposes in GIP-2 to use 5% damped seismic demand spectra for 
   comparison with the corresponding seismic capacity spectra.  The staff 
   has examined the damping values listed in the licensing basis documents 
   of the USI A-46 plants.  Several of these plants have been licensed with 
   equipment damping values of 2% or less.  However, the majority of the 
   plants in this group do not have a clear definition of the damping values 
   for equipment in their safety analysis reports (SARs).  The seismic 
   capacity spectra of the equipment in the seismic experience database were 
   established at 5% damping.  Although in the amplified range at discrete 
   frequencies the seismic demand spectrum for equipment at 2% is higher 
   than that at 5%, the seismic capacity spectrum at 2% is also higher than 
   that at 5%.  The difference in magnitude (in the ordinates) between the 
   seismic demand spectra at 2% and 5% is comparable to the difference in 
   magnitude between the seismic capacity spectra at 2% and 5%.  For the 
   purpose of comparison, insofar as verifying seismic adequacy of equipment 
   using the GIP-2 methodology, it is the judgement of the staff that the 
   damping level at which the seismic demand spectra are established is of 
   little significance (for the range of damping values discussed herein, 
   i.e. approximately 1-5%) provided that the 
   .   

   corresponding capacity spectra are established at the same damping 
   levels.  Therefore, the staff finds that the use of seismic demand 
   spectra for comparison at 5% damping is acceptable for all USI A-46 
   plants for the purpose of verifying the seismic adequacy of equipment.

3. With respect to the "Definition of Terms" in the last paragraph of page 
   4-18 of GIP-2, the staff positions on the definition of "conservative, 
   design" in-structure response spectra are as follows: "Conservative, de-
   sign" in-structure response spectra are defined as in-structure response 
   spectra that have been computed in accordance with current NRC regulatory 
   guidelines (such as RG 1.60 and RG 1.61) and the Standard Review Plan 
   (SRP Section 3.7, Rev. 2, August 1989).  Alternatively, for post-1976 
   operating license (OL) plants with non-Housner-type ground response 
   spectra (Category 1 plants without double asterisks , Table A) and plants 
   included in the Systematic Evaluation Program (SEP, Category 2, Table A), 
   the in-structure response spectra included in the licensing-basis (LB) 
   documents such as final safety analysis reports (FSARs), updated safety 
   analysis reports (USARs), and other pertinent commitments related to 
   in-structure response spectra may be used as "conservative, design" 
   in-structure response spectra.  For plants in neither category 
   (Category 1 plants with double asterisks and Category 3, Table A), the 
   plant LB in-structure response spectra may be used, provided that the 
   licensee submits as part of its 120-day response package the detailed 
   information on which procedures and criteria were used to generate those 
   in-structure response spectra (see item 5, Section 2.2.1 of Part I of 
   GIP-2).  The staff will review the acceptability of the proposed usage 
   case-by-case.  The staff approval of the proposed in-structure response 
   spectra is necessary before the commencement of the implementation 
   program.

   As stated in Section 2.2.1 of Part I of GIP-2, each licensee shall submit 
   its schedule for implementing the resolution of USI A-46 within 120 days 
   after this supplement is issued.  The plant-specific implementation 
   schedule shall be such that the affected plant should complete its 
   implementation within 3 years after the issuance of this supplement.  For 
   Category 1 plants with double asterisks and Category 3 plants, however, 
   the 3-year period will not commence until one of the following conditions 
   is met:

   (1)      the receipt of staff approval of the in-structure             
            response spectra to be used to resolve the USI A-46.        

   (2)      60 days following the licensee's initial submittal of acceptable 
            procedures and criteria in generating those in-structure 
            response spectra.




                                  Table A  
                  USI A-46 plants categorized according to
                       In-structure Response Spectra* 
 
 Category 1             Category 2            Category 3

Post-1976 OL            SEP plants            Pre-1976 OL plants 
plants    

(15 units)              (9 units)              (40 units)

  Arkansas 2            Palisades              Robinson 2     
**Crystal River 3       Ginna                  Point Beach 1/2  
**St. Lucie 1           Oyster Creek           Monticello  
  Hatch 2               Dresden 2              Dresden 3   
**Calvert Cliffs 2      Millstone Unit 1       Pilgrim 1    
**Cook 2             ***Yankee Rowe            Quad Cities 1/2 
**Salem 1/2             Haddam Neck            Surry 1/2  
  Brunswick 1           Big Rock Point         Turkey Point 3/4
  Davis-Besse 1         San Onofre 1           Oconee 1/2/3  
  Beaver Valley 1                              Brunswick 2
  North Anna 1/2                               Trojan 
**Browns Ferry 3                               Millstone 2
  Farley 1                                     Nine Mile Point 1
                                               Peach Bottom 2/3
                                               Prairie Island 1/2 
                                               Duane Arnold
                                               Cooper
                                               Arkansas 1
                                               Calvert Cliffs 1
                                               Cook 1
                                               Hatch 1
                                               FitzPatrick
                                               Three Mile Island 1 
                                               Vermont Yankee
                                               Kewaunee
                                               Fort Calhoun
                                               Zion 1/2
                                               Browns Ferry 1/2
                                               Indian Point 2/3 



*     All plants in this table are the USI A-46 plants.  All plants except 
      St. Lucie 1 and Turkey Points 3/4 are SQUG members.  In case more than 
      one set of in-structure response spectra appear in the LB documents, 
      use the more conservative set of spectra or justify the use of the 
      others.

**    Category 1 Plants with Housner-type ground response spectra.

***   Yankee Rowe is no longer an operating reactor.

.

4. Regarding the scaling of the Individual Plant Examination of External 
   Events (IPEEE) spectra for application in the USI A-46 program (page 4-19 
   of GIP-2), the staff's position is as follows:

   The in-structure response spectra for some USI A-46 plants may have been 
   or may be developed for the IPEEE based on the realistic, median-center 
   method as described in Section 4.2.4 of Part II of GIP-2.  This method 
   uses the NUREG/CR-0098, "Development of Criteria for Seismic Review of 
   Selected Nuclear Power Plants," 1978, median rock or soil spectrum 
   (depending on the primary condition at the site) anchored at the assigned 
   review level earthquake.  For these plants, the IPEEE in-structure 
   response spectra may be used to generate realistic, median-centered 
   in-structure response spectra for use in the USI A-46 program by 
   appropriately scaling down the IPEEE spectra. 

   If this approach is to be used to resolve USI A-46, the licensee should 
   submit as part of its 120-day response package the procedure and the 
   criteria to be used to generate those realistic, median-centered in-
   structure response spectra if different than those specified in GIP-2 
   (see Item 5, Section 2.2.1 of Part I of GIP-2).  But, when technical 
   justifications identified on page 4-19, Item d and the last sentence of 
   Page 4-20 of GIP-2 need to be developed, then such justifications should 
   be provided to the NRC in the 120-day response package.

   This staff position is intended, for USI A-46, to allow the licensee to 
   use seismic input for the IPEEE as described in NUREG-1407, "Procedural 
   and Submittal Guidance for the Individual Plant Examination of External 
   Events (IPEEE) for Severe Accident Vulnerabilities:  Final Report," June 
   1991.

II.4.3  Equipment Class Similarity and Caveats

The staff interprets Section 4.3 of Part II (as well as Sections 4.1.3 and 
4.2.2) regarding the use of caveats to mean that the review engineer will 
determine whether the equipment meets both the caveats and their intent and 
will report accordingly (i.e., via the Seismic Evaluation Report to be 
submitted to the NRC and in Appendix G of the GIP).

II.4.4  Anchorage Adequacy

Regarding anchorage guidelines, GIP-2 provides criteria and screening 
procedures for five types of anchorages that are used to secure an item or 
equipment:  (1) expansion anchors, (2) cast-in-place bolts and headed studs, 
(3) cast-in-place J-bolts, (4) grouted-in-place bolts, and (5) welds to 
embedded or exposed steel. GIP-2 classifies any other types of anchorage as 
outliers.

1.   Expansion Anchors

   For expansion anchors, GIP-2 provides nominal allowable pullout and shear 
   loads for various diameters of single anchors for certain concrete 
   strengths with specified minimum embedments, minimum spacings between 
   .   

   anchors, and minimum distances of anchors to a free concrete edge.  Also 
   provided in GIP-2 are load reduction factors for specific types of 
   expansion anchors by different manufacturers.  GIP-2 requires a tightness 
   check on the anchor head or nut to detect gross installation defects. 

   Acceptance criteria are provided to assure a 95-percent confidence level 
   that there are no more than 5-percent anchors which do not meet the 
   tightness check, as described in GIP-2, for expansion anchors.  GIP-2 
   also requires a check on the anchor projection above concrete to ensure a 
   minimum anchor embedment in the concrete.  Furthermore, checks are 
   required on the spacing between anchors, the distances from anchors to a 
   free concrete edge, concrete strength, and concrete cracking conditions, 
   and reduction factors for nominal allowable loads are specified in the 
   GIP for each condition which does not meet the minimum requirements for 
   anchors having nominal allowable loads.  On the basis of this 
   information, the actual allowable load for each anchor can be calculated.

   The nominal allowable pullout and shear loads for single anchors in GIP-2 
   were obtained by dividing the average ultimate loads of test expansion 
   anchors by a minimum safety factor of three.  The reduction factors 
   specified in GIP-2 for different manufacturers, for closely spaced 
   anchors, for less edge distance than the specified minimum, for less 
   concrete strength than that of anchors having nominal loads, and for 
   cracked concrete were also obtained from test data.  The staff concludes 
   that a minimum safety factor of three in conjunction with appropriate 
   reduction factors for other conditions as specified in GIP-2 gives 
   adequate safety margins for allowable loads of expansion anchors.  
   Furthermore, the staff concludes that the safety margin of expansion 
   anchors is enhanced by the GIP-2 requirements of 100-percent visual 
   inspection of accessible anchors and sample tightness checks of expansion 
   anchors.

2. Cast-in-Place Bolts and Headed Studs

   As with expansion anchors, GIP-2 provides nominal allowable pullout and 
   shear loads for various diameters of single bolts and studs for certain 
   concrete strengths with specified minimum embedment, minimum spacings 
   between bolts or studs, and minimum distances of bolts or studs to a free 
   concrete edge.  GIP-2 requires a check on the actual embedment, spacing 
   between bolts and studs, distances of bolts or studs to a free concrete 
   edge, concrete strength, and concrete cracking conditions for 
   cast-in-place bolts and headed studs, and specifies reduction factors for 
   allowable loads for each condition which does not meet the minimum 
   requirements for bolts and studs having nominal allowable loads.  On this 
   basis, the actual allowable load for each bolt or stud can be calculated.

   The nominal allowable shear loads for single bolts are based on the 
   nominal bolt area times allowable shear stress of 17 ksi.  The staff 
   compared the allowable shear loads of single bolts specified in GIP-2 
   .   

   with test data, and found that safety factors with respect to ultimate 
   failure loads are greater than three, which is adequate.  The nominal 
   allowable pullout loads for single bolts or studs are based on the 
   nominal bolt area times allowable tensile stress of 34 ksi, and have a 
   safety factor of two with respect to a 45-degree concrete cone failure 
   mechanism.  For anchorages with multiple bolts or studs, a minimum safety 
   factor of one and one-half is provided against a 45-degree concrete cone 
   failure.  The 45-degree failure cone was also assumed for the effects of 
   bolts spaced close to each other or located close to the concrete free 
   edge.  The 45-degree failure cone shape is hypothetical, and predicts a 
   lower pullout load for bolts with shallow embedment and a higher pullout 
   load for bolts with deep embedment than that of test bolts.  The staff 
   has verified that, even for the deepest embedded bolt presented in GIP-2, 
   the actual safety factor provided by GIP-2 is still slightly greater than 
   one and, therefore, the staff accepts the allowable loads as specified in 
   GIP-2.  The reduction factor given in GIP-2 for cracked concrete was 
   based on test data and, thus, is also acceptable to the staff.

3. Cast-in-Place J-Bolts

   A J-bolt is a plain steel bar that has a hook (usually in a 90- or 180-
   degree form) at the end which is embedded in concrete, and is threaded 
   with a nut at the other end.  GIP-2 provides nominal allowable pullout 
   and shear loads for various diameters of single J-bolts for concrete 
   strength equal to or greater than 3.5 ksi with specified minimum 
   embedments, minimum spacings between bolts, and minimum distances of 
   bolts to a free concrete edge.  GIP-2 specifies reduction factors for 
   bolts that are embedded less than the specified minimum, that are located 
   closer to a free concrete edge than the specified minimum, and that are 
   embedded in concrete with strength less than 3.5 ksi.  GIP-2 requires 
   that J-bolts be classified as outliers if the bolts are spaced less than 
   three times the bolt diameter, or if they are embedded in certain cracked 
   concrete.

   The nominal allowable pullout and shear loads for single J-bolts are 
   identical to that of cast-in-place bolts and headed studs.  Since the J-
   bolts are embedded in concrete much deeper than the cast-in-place bolts 
   or headed studs, the J-bolts can only fail either in the steel material 
   or if the J-bolt is pulled out upon failure of the concrete bond.  The 
   specified minimum embedments for J-bolts provide a safety factor of about 
   two with respect to concrete bond failure, which is acceptable to the 
   staff.  The allowable shear loads have safety factors greater than three 
   with respect to ultimate shear failure loads, which is also acceptable to 
   the staff.  The reduction factor in GIP-2 for pullout is in proportion to 
   the reduction in the straight portion of J-bolt embedment.  This is 
   reasonable because bond force from concrete to bolts is proportional to 
   the embedment length and, therefore, is acceptable to the staff.  The 
   reduction factor in GIP-2 for pullout and shear loads due to concrete 
   strength less than 3.5 ksi is proportional to the square root of the 
   ratio of the actual strength to the nominal strength of 3.5 ksi.  This is 
   also reasonable because this reduction represents concrete 
   .   

   tensile strength reduction, and thus reduces the holding power of bolts.  
   Therefore, the staff concludes that the use of appropriate safety factors 
   for single J-bolts in conjunction with appropriate reduction factors 
   applied to various conditions as specified in GIP-2, would provide 
   adequate safety margins for allowable loads of cast-in-place J-bolts.

4. Grouted-in-Place Bolts

   GIP-2 provides nominal allowable pullout and shear loads for various 
   diameters of single, grouted-in-place bolts for concrete strength equal 
   to or greater than 3.5 ksi with specified minimum spacings between bolts, 
   and minimum distances of bolts to a free concrete edge.  GIP-2 requires a 
   check on the actual embedment, spacing between bolts, distance of bolts 
   to a free concrete edge, concrete strength, and concrete cracking 
   conditions for grouted-in-place bolts, and specified reduction factors 
   for allowable loads for each condition that does not meet the minimum 
   requirements for bolts having nominal allowable loads.  On this basis, 
   the actual allowable load for each bolt can be calculated.

   The provisions for grouted-in-place bolts in GIP-2 are identical to 
   provisions for cast-in-place bolts and headed studs if bolts were found 
   to be installed using certain installation procedures.  However, if such 
   installation procedures cannot be verified to have been used, GIP-2       
   reduces the nominal allowable pullout loads to one-tenth of that of 
   cast-in-place bolts and headed studs with other provisions remaining 
   unchanged.

   Test results have indicated that grouted-in-place-bolts, installed 
   properly, can develop the same allowable loads as cast-in-place bolts.  
   However, test results also show that pullout loads of grouted-in-place-
   bolts drop substantially if the bolts were not installed properly.  The 
   staff believes that the use of 10 percent of the allowable loads (as 
   specified in GIP-2) of properly installed grouted-in-place-bolts for the 
   bolts for which proper installation procedures cannot be verified is 
   conservative.  The staff also believes that the allowable shear loads and 
   other phenomena of grouted-in-place bolts should be similar to that of 
   cast-in-place bolts and headed studs.  Therefore, the staff concludes 
   that the provisions in GIP-2 for grouted-in-place bolts are adequate.

5. Welds to Embedded or Exposed Steel

   GIP-2 provides allowable loads for welds of various sizes, and requires 
   an inspection of the weld size and quality.  The minimum effective length 
   of fillet welds should not be less than four times the nominal size of 
   the weld, or else the size of the weld should be considered not to exceed 
   one-fourth of its effective length.  The allowable loads are based on the 
   weld size times an allowable weld stress of 30.6 ksi.  The staff 
   concludes that the allowable loads so determined for such weld 
   calculations are conservative and provide adequate safety margins against 
   failure for welds to embedded or exposed steel.
.

6. Determination of Seismic Load for Individual Anchor

   GIP-2 states that the seismic load on anchorages can be calculated by 
   assuming an equivalent static load acting on the center of gravity of the 
   equipment, with the load being equal to the input seismic accelerations 
   times the mass of the equipment.  GIP-2 further states that the seismic 
   accelerations can be obtained from any one of the following three types 
   of response spectra:  (1) a "conservative, design" horizontal, 
   in-structure response spectrum for SSE as defined in GIP-2, and modified 
   by Item 3 of Section II.4.2 of this supplement with no modification 
   factor, (2) a median-centered, horizontal, in-structure response spectrum 
   for SSE as defined in GIP-2, and (3) a 1.5 times SSE horizontal ground 
   response spectrum for equipment mounted 40 feet above the effective grade 
   and having its lowest natural frequency greater than 8 Hz.  If option (2) 
   or (3) is selected, the acceleration is increased by a modification 
   factor of 1.25.  The vertical component of acceleration is assumed to be 
   two-thirds of the horizontal component of acceleration.  The 
   square-root-of-the-sum-of-the-squares (SRSS) method is used to combine 
   the load components from three directional accelerations.  The final load 
   on each anchor is calculated by adding the combined seismic loads to the 
   equipment deadweight loads and any other loads on the anchor.  The staff 
   concludes that the procedures specified in the GIP for determining loads 
   on individual anchors provide adequate safety margin against failure and 
   are, therefore, acceptable.

7. Modification and Replacement of Expansion Anchors

   GIP-2 states the following:

      The GIP-2 criteria may be applied to modification or repair of 
      existing anchorages (e.g., anchor bolts or welds) including one-
      for-one component replacements (e.g., replacing bolts in one-for-one 
      component replacements). 

      For new installations and newly designed anchorages in modifications 
      or replacements, the GIP-2 criteria and procedures may also be 
      applied, except that the factor of safety currently recommended for 
      new nuclear power plants in determining the allowable anchorage loads 
      shall be met.

      It is generally recommended that if expansion anchors need to be used 
      for vibrating equipment, then the undercut type of expansion anchors 
      should be installed.

   The staff concurs with these statements because they are practical and 
   reasonably conservative.

8. Identification and Resolution of Outliers

   Anchors are classified as outliers if the loads acting on the anchors 
   exceed their allowable loads, or if anchors fail to pass certain 
   screening guidelines specified in GIP-2.  GIP-2 requires the licensee to 
   .   

   assign qualified persons to the task of outlier resolution.  Although 
   GIP-2 provides recommendations on generic methods for resolving outliers, 
   it states that the details for resolving outliers are beyond its scope.  
   GIP-2 further states that the utility is responsible for resolving 
   outliers using its existing engineering procedures as it would resolve 
   any other seismic concern.  The staff considers the task of outlier 
   resolution to be plant specific, and agrees that the acceptability of the 
   outlier resolution should be addressed individually by each licensee.  
   The staff will, in its plant-specific SER, present its evaluation of the 
   licensee's proposed method for resolving any outliers identified in the 
   plant-specific walkdown inspection summary reports.

9. Verification of Anchorage Capacity by Computer Codes

   Two computer codes, EBAC code (Reference 7 of GIP-2) and ANCHOR code 
   (Reference 14 of GIP-2), were developed and referenced in the GIP-2 for 
   verifying anchorage capacity of equipment in USI A-46 plants.  

   The EBAC 1.0 and ANCHOR 3.0 anchorage evaluation computer codes use 
   somewhat different analysis approaches to determine the adequacy of a 
   given anchorage arrangement and postulated seismic loading.  Although 
   both codes use a seismic equivalent static load approach to evaluate the 
   equipment anchorages, the application of the equivalent static loads 
   differs between these two codes.  The EBAC code applies the seismic 
   equivalent static loads in one direction at a time to an anchorage and 
   then takes the square root of the sum of the squares (SRSS) of the bolt 
   reactions from the three-directional seismic inputs.  The ANCHOR code 
   allows an SRSS combination of three-directional seismic equivalent static 
   loads applied simultaneously and then the combined load is applied at the 
   equipment center of gravity.  The staff finds both methodologies for load 
   application acceptable.  

   A major difference between these two computer codes was noted in the 
   selection of the equipment overturning axis (i.e., the neutral axis).  
   The EBAC code performs a linear elastic analysis in assuming that plane 
   sections remain plane as the overturning moments are applied.  This 
   assumption leads to a linear distribution of the tension forces on the 
   anchors due to an overturning moment.  No compressive forces in the 
   concrete are considered.  The EBAC code asks the user to input the 
   equipment overturning axis locations based on perceived equipment base 
   flexibility.  The staff finds that the lack of specificity regarding the 
   location of the overturning axis could lead to underestimation of anchor 
   loads.  Furthermore, concrete crushing strain limit could be exceeded if 
   they are not verified by the user.  In contrast, the ANCHOR code uses an 
   approach which assumes an anchorage to exhibit an elastic-perfectly-
   plastic behavior for the anchor bolts and the concrete.  The ANCHOR code 
   computes the overturning axis based on the overturning moments being 
   resisted by tensile forces in the anchors and compressive forces in the 
   concrete, assuming the base plate to be rigid.  The user should consider 
   the limit for concrete crushing strain (strength) and verify the 
   applicability of the rigid base plate assumption.
.

   The EBAC code provides for a bilinear and an exponential tension/shear 
   interaction formulation for anchor bolt strength evaluation.  The ANCHOR 
   code allows selection of tension/shear interaction factors to represent a 
   bilinear interaction formulation.  For both computer codes, the selection 
   of tension/shear interaction formulation must be consistent with those 
   given in Appendix C of GIP-2.
 
   Therefore, the EBAC code as given in Reference 7 of GIP-2 and the ANCHOR 
   code as given in Reference 14 of GIP-2 are acceptable provided that the 
   items of concern discussed above are adequately considered prior to their 
   applications. Other computer codes may be used for anchorage evaluations 
   if demonstrated to be acceptable.

   Furthermore, for overall anchorage design and analysis, the equipment 
   anchorage attributes listed in Section 4.4.1 of Part II of GIP-2 and the 
   concerns described on page 49 of Reference 5 in GIP-2 must also be taken 
   into consideration.  

10.   Anchorages in Inaccessible Areas

   Regarding the verification of anchorages in inaccessible areas, GIP-2, 
   states on page 4-28 that "inaccessible anchorages not required for 
   strength...need not be inspected...."  This is, in general, acceptable to 
   the staff because, if the inaccessible anchorages are not required for 
   strength, it implies that the structural integrity of the anchorage is 
   already adequate.  However, to ensure the relay functionality, the 
   licensee should try all practicable means to inspect all the anchorages 
   of the cabinets having essential relays to avoid impact or excessive 
   cabinet motion.

11.   Minimum Spacing Between Anchors

   The sentence "The minimum spacings given in Appendix C are for distances 
   between adjacent anchors in which the cones of influence just touch each 
   other at the surface of the concrete...."  which starts at the end of 
   page 4-39 in GIP-2, is incorrect.  This is because the values of minimum 
   spacing given in Appendix C of GIP-2 were directly taken from Volume 1 of 
   GIP-2 Reference 7, and these values correspond to a 13-percent shear cone 
   overlapping as stated on page 2-81 of Volume 1 of GIP-2 Reference 7.  
   Therefore, this quoted statement should be corrected to be consistent 
   with the statement given in GIP-2 Reference 7.

12.   Use of ACI 349

   On page 4-32, GIP-2 states that any anchorage other than the five types 
   of anchorages covered in GIP-2 should be classified as an outlier and be 
   resolved in accordance with the guidelines described in Section 5 of 
   GIP-2.  Therefore, other types of cast-in-place embedments as stated in 
   "Check 14 - Embedment Steel and Pads" in Section 4.4.1 of GIP-2 should be 
   treated as outliers and resolved in accordance with the guidelines 
   described in Section 5 of GIP-2.  The proposed ACI-349 Code method of 
   resolving these special outliers is beyond the scope of GIP-2 and also, 
   .   

   the entire second paragraph except for the first two sentences on page 
   4-49 of GIP-2 is not acceptable to the staff.  The licensee should use 
   Appendix C to GIP-2 for guidance.

13.   Frequency Shifting

   On pages 4-10 (fourth paragraph) and 4-56 (2nd from last paragraph), when 
   an unbroadened seismic demand response spectrum is used for comparison, a 
   reference or basis should be provided by the licensee for methods of 
   "frequency shifting" for addressing the uncertainty in natural frequency 
   of the building structure. 

II.5  Outlier Identification and Resolution

Discussion

Section 5 of Part II defines an outlier as an item of equipment that does 
not meet the screening guidelines provided in GIP-2.  Several generic 
methods for resolving outliers are summarized in Section 5 of Part II.

Evaluation

As noted in Section 5.3 of Part II, the details for resolving outliers are 
beyond the scope of the GIP.  It is the responsibility of the utility to 
resolve outliers, using existing procedures (e.g., plant-specific procedural 
controls and QA requirements) as it would resolve any other seismic 
concerns.  Therefore, the methods and results of outlier resolutions will be 
treated on a plant-specific basis.

It should be noted that one of the methods suggested in GIP-2 for resolving 
outliers is to use the earthquake experience data documented in References 4 
and 5 of GIP-2.  However, GIP-2 Reference 4 has been modified extensively by 
the addition of unreviewed new information.  Although the staff and GIP-2 
Reference 5 (SSRAP report) have used the information contained in a previous 
draft version (dated February 1987) of GIP-2 Reference 4 to assist the staff 
in arriving at a decision for resolution of some technical issues, the staff 
has not reviewed this GIP-2 Reference 4, and the GIP-2 Reference 5 does not 
endorse the entire GIP-2 Reference 4 (see Reference 9).  Therefore, any 
specific application of the detailed information documented in GIP-2 
Reference 4 for the implementation of USI A-46 resolution should be 
submitted to the NRC staff for review and approval before it is used.  
Regarding the evaluation of the acceptability of new data, the staff 
position is described in Section I.3.0 of this supplement. 


Conclusion

Subject to the above clarifications, the staff concludes that the procedures 
for outlier identification and the general approach for outlier resolution 
are adequate and acceptable.  

.

II.6  Relay Functionality Review

Discussion

Section 6 of Part II provides an overview of the relay evaluation procedure 
and describes the relationships between other GIP activities and the relay 
evaluation which is contained in a separate reference document, "Procedure 
for Evaluating Nuclear Power Relay Seismic Functionality," Reference 8 of 
GIP-2.

This section was revised from GIP-0, primarily to include a multilevel 
screening approach for comparing relay seismic capacity to demand.  The 
evaluation procedure described in GIP-2 is a summary; the details of the 
method are contained in the above-referenced SQUG relay procedures.

Evaluation and Conclusion

The relay review requires the use of the generic equipment ruggedness 
spectra (GERS) to assess the relay ruggedness.  The staff had concerns about 
the amount of relay data that were available in the GERS (open issue C.2.1 
of the original GSER).  Additional information has been considered in the 
GERS and data will be added as needed if the walkdowns identify relays not 
currently addressed.  The staff considers the SQUG approach practical and, 
therefore, acceptable and this issue resolved.

The GERS were constructed using test data from relays of vintages newer than 
those that are currently installed in the USI A-46 plants.  In the GSER, 
open issues C.2.2 and E.2.2 described the concern that the testing of newer 
equipment may not be applicable to the older equipment.  The SQUG initiated 
a program to test a sample of older relays which were of the same type as 
those covered by the GERS, and compared the results to the more recent test 
results.  The test results demonstrated that the difference in seismic 
ruggedness between relays of different vintages was not significant.  GIP-2 
considers this issue resolved, and the staff concurs.  If additional testing 
in the future, by the NRC, SQUG, or others, provides evidence to change this 
conclusion, the staff will take appropriate action at that time. 

Open issue E.2.5 of the GSER discussed the inclusion of relay mountings in 
the walkdown inspection and the number of relays to be inspected.  The staff 
concurs with the SQUG position to review a sample of the relay mountings to 
ascertain that the relays are mounted in conformance with the vendors' 
recommendations.  If any abnormality exists, the licensee shall increase the 
number of samples for inspection.

In conclusion, on the basis of its review of Section 6 of Part II, the staff 
agrees with the approach of evaluating systems and electrical circuits to 
determine the effect of relay chatter and endorses the review procedure as 
given in GIP-2.  Therefore, the staff concludes that the procedure, if 
properly implemented, is an acceptable method of verifying the seismic 
adequacy of relays for the resolution of USI A-46 subject to the following:

.

1. Use of Zero Period Acceleration Capacities

   Regarding the acceptability of a relay, because of the important effects 
   of zero period acceleration (ZPA) on relay chatter, the staff position is 
   that in addition to the comparison of the spectral accelerations, the ZPA 
   capacities should be compared and shown to be adequate (page 6-18 of 
   GIP-2.)

2. Development of In-Cabinet Amplification Factors

   Section 6 of Part II includes the use of a single number amplification 
   factor which is applicable to a given class of equipment for what is 
   defined as Screening Level 2.  With this concept, an in-cabinet demand 
   spectrum is estimated by multiplying the base excitation demand spectrum 
   by an effective amplification factor that is representative of the given 
   class of equipment.  The result is then compared with device ruggedness 
   spectra to verify device capability.

   The amplification factors for motor-control-center-type cabinets, for 
   control room benchboards and panels, and for switchgear-type cabinets or 
   similar panels are presented in Table 6-2 of GIP-2.  Because these 
   amplification factors were determined based on test data and some 
   empirical parameters specific to a certain type of cabinets or panels, 
   the staff concludes that these amplification factors are reasonable and 
   acceptable.  However, the use of the 0.6 reduction factor for narrow peak 
   amplification spectra for other types of cabinets, panels, or enclosures 
   must be justified by the user and documented using procedures described 
   in Reference 2 of Section 4 of GIP-2 Reference 33 because this 0.6 factor 
   is an empirical value derived from specific types of cabinets, panels, or 
   enclosures.

3. Development of In-Cabinet Response Spectra

   The use of the in-cabinet amplification factors is intended for initial 
   screening purposes.  Should the result not produce a positive equipment 
   seismic verification in a given case, then the next level of screening 
   presents a more definitive methodology developed by EPRI (GIP-2 Reference 
   33) for generating in-cabinet response spectra.  The staff has reviewed 
   the procedures described in GIP-2 Reference 33 and the results 
   specifically applicable to control room benchboards and panels.  The 
   staff finds that the approach includes development of conservative 
   estimates for a single generic lowest natural frequency and a 
   corresponding single generic high participation factor for this class of 
   equipment.  Therefore, the staff concludes that GIP-2 Reference 33 
   constitutes an approximate method of generating an in-cabinet demand 
   response spectrum for devices which will be attached to control room 
   benchboards and panels that are subject to a given site-specific floor 
   spectrum.

   The EPRI methodology (GIP-2 Reference 33) includes a combination of in 
   situ experimental tests, modal analysis, linear power spectral density 
   (PSD) response prediction, response spectrum/PSD transformation, and 
   .   

   statistical methods in various combinations to generate the final, 
   generic, elevated-demand spectrum.  The staff finds the use of the 
   computer program GENRS, as documented in Reference 33 of GIP-2 for the 
   calculation of in-cabinet response spectra, acceptable only for control 
   room benchboards and panels as defined in GIP-2 Reference 33 because the 
   parametric values, such as those for natural frequency and the 
   corresponding participation factor used in the computer code (GENRS), 
   were derived specifically from the control room benchboards and panels.  
   Therefore, the use of GENRS should not be extended to other classes of 
   equipment without the review and approval of the NRC staff.

   The EPRI methodology includes direct generation of a PSD from a required 
   response spectrum (RRS) and vice versa.  Because, the current NRC staff 
   position on this approach is that the direct-generation method can be 
   considered only case by case, the staff performed some additional 
   investigation concerning the viability of this approach and its 
   applicability in the GENRS computer code.  The results of the staff 
   investigation support the viability of the direct-generation method in 
   general and its application in the GENRS computer code in particular.

II.7  Tanks and Heat Exchangers Review

Discussion

Section 7 of Part II gives guidelines for evaluating the adequacy of tanks 
and heat exchangers.  The SQUG commitments, evaluation methodology, vertical 
tanks, horizontal tanks, outliers, and documentation are the main topics in 
this section. 

Evaluation

1. Vertical Tanks

   The procedure given in Section 7 of Part II of GIP-2 and discussed in the 
   subsequent paragraphs covers the screening guidelines for flat-bottom 
   vertical tanks supported on a concrete pad or floor, and anchored to the 
   pad (or floor) by means of cast-in-place anchor bolts.  The screening 
   guidelines are applicable when the tank dimensions, anchor bolt 
   configurations, and materials of fabrication are within the range and 
   assumptions given in Table 7-1 in Part II.

   The last paragraph in Section 7 of Part II indicates that the successful 
   completion of the review described in Section 7 has been accepted by the 
   NRC as resolving the seismic issues related to these types of tanks for 
   USI A-40.  However, the SQUG commitments in Section 7.1 and evaluation 
   methodology in Section 7.2 do not address the screening guidelines for 
   ensuring the adequacy of the foundation structures of the vertical tanks.  
   As the tank foundation is subjected to higher loads than those determined 
   using the rigid tank assumption, SRP Section 3.7.3.II.14.i recommends 
   that the tank foundation be designed to withstand the seismic forces 
   imposed on it.  The SQUG commitments in Section 7.1 are not consistent 
   with the guidelines in SRP Section 3.7.3.II.14.i.  However, 
   .   

   the resolution to this issue as discussed in the later part of this 
   section is acceptable to the staff for tanks in USI A-46 plants.  

   Because the screening guidelines are to be used for the as-built vertical 
   tanks, the staff strongly recommends that the input data required in Step 
   1 of Section 7.3.2 be based on the pertinent as-built drawings and 
   verification through walkdowns of the condition of the tanks and the 
   supporting foundations.  Steps 2 through 6 provide the guidelines for 
   determining the seismic demand applied to a specific tank, in terms of 
   the overturning moment and the shear load.  The seismic demand is based 
   on the response value of the fluid-structure model at the impulsive modal 
   frequency (Step 4).  The calculated frequency is varied by � 20 percent 
   to account for the uncertainties involved in the calculations.  The 
   maximum responses from the applicable ground or floor response spectrum 
   at 4-percent damping are used to calculate the seismic demand.  
   Guidelines are provided to account for soil-structure interaction effects 
   on the frequency and the response.  On the basis of its review of 
   procedures described in Steps 1 through 6 of Section 7.3.2 of Part II as 
   summarized above, the staff finds that the seismic demand so determined 
   is adequate and, therefore, concludes that these steps are logical and 
   acceptable.

   Steps 7 through 18 of Section 7.3.2 of Part II provide a method for 
   computing the overturning moment capacity of the tank.  The method 
   considers the complex interactions between the anchor bolt capacity, the 
   anchorage connection capacity, and the allowable buckling stress.  Steps 
   19 and 20 require the users to compute the shear load capacity provided 
   by the weight of the fluid on the base of the tank, and the frictional 
   resistance between the base of the tank and the foundation surface.  The 
   formula to compute shear load capacity also reduces the fluid weight to 
   account for the 40 percent of the vertical component of the earthquake.  
   Steps 21 and 22 require the users to evaluate fluid level against the 
   slosh height computed for the postulated earthquake.  Section 7.3.6 of 
   Part II requires the users to check the effect of the flexibility of 
   attached piping. 

   In reviewing the earlier version of GIP-2 (Reference 6), the staff 
   identified the following concerns:

   a. The SQUG commitments do not require the users to check the adequacy of 
      the supporting foundation which are likely to be subjected to higher 
      loads than the original design that was based on a rigid-tank 
      assumption.

   b. The allowable buckling stress criteria provided in Step 11 are not 
      sufficiently conservative to account for the out-of-roundness of the 
      tank, local imperfections, material nonlinearities, the secondary 
      effects due to shearing stresses, and rotation of the shell wall at 
      the base.  Without considering the uncertainties induced by these 
      inherent characteristics, the seismic adequacy of the tanks cannot be 
      assured.

.

   In order to resolve the concern regarding the adequacy of the tank 
   foundation, the SQUG proposed to include the evaluation requirements for 
   ring foundations of the vertical tanks.  The SQUG justified the narrow 
   scope of the requirements by pointing to the experience data regarding 
   tank foundation failures.

   The staff agrees that ring foundations, when subjected to loads higher 
   than the design loads, are likely to be more susceptible to failure than 
   other types of foundation such as foundation mats on ground or floors 
   supporting the tanks.  Therefore, the proposed resolution is acceptable 
   to the staff, and the concern is resolved with the inclusion of 
   instructions to the users in Section 7.3.7 of Part II to identify ring 
   foundations as outliers.  

   In order to resolve the concern regarding allowable buckling stress 
   capacity, the staff has proposed to reduce the capacity reduction factor 
   in Step 16 of Section 7.3.2 Part II from 0.9 to 0.72.  The SQUG has 
   adopted the staff recommendation in GIP-2.  Therefore, this concern is 
   resolved.

   During the discussion related to the resolution of USI A-40, "Seismic 
   Design Criteria," the method of analysis of above-ground, flexible, 
   vertical tanks was identified as a topic requiring technical resolution.  
   USI A-40 is resolved in Standard Review Plan (SRP) Sections 2.5.2, 3.7.1, 
   3.7.2, and 3.7.3 (Revision 2, August 1989).  The guidelines related to 
   the seismic analysis of the above-ground vertical tanks are included in 
   SRP Section 3.7.3.II.14.  As part of the resolution of USI A-40, a number 
   of tanks at nuclear power plant sites are required to have confirmatory 
   checks to ensure that the safety-related, above-ground, vertical tanks 
   are adequately designed.  Most of the licensees of newer plants have 
   incorporated the flexible tank design concept in the design of their 
   above-ground tanks.  Some licensees have committed to make confirmatory 
   checks of their design using the procedures developed by the SQUG under 
   the resolution of the USI A-46 program.  The implementation of criteria 
   and procedures described in GIP-2, supplemented by the staff evaluations 
   described in this supplement for large, flat-bottom, cylindrical, 
   vertical tanks which are needed for safe shutdown and for refueling water 
   storage in PWRs, is considered an acceptable method for resolving the 
   seismic issues related to these types of tanks for both USI A-46 and USI 
   A-40, as it applies to USI A-46 plants. 

2. Horizontal Tanks

   The screening guidelines provided in Section 7.4 of Part II of GIP-2 are 
   applicable when a horizontal tank or a heat exchanger shell satisfies the 
   following criteria: 

      Its longitudinal axis (axis of symmetry) is horizontal.

      It is supported on its curved bottom by steel saddle plates.

.

      It is anchored to a stiff foundation having adequate strength to 
      resist the seismic loads applied to the tank.

      All the baseplates under the saddle have slotted anchor-bolt holes in 
      the longitudinal direction except the one for an end saddle support. 

      Its layout and dimensions satisfy the range of parameters and 
      assumptions listed in Table 7-6 of Part II. 


   Steps 2 through 7 of the screening guidelines described in Section 7.4 of 
   Part II provide guidelines for evaluating the resistance of the existing 
   tank in terms of the anchorage capacity.  Steps 8 through 10 provide 
   guidelines for evaluating the seismic demand of the tank anchorage 
   system.  Step 11 provides instructions for evaluating the tank saddle 
   stresses.  The staff finds the screening methodology to evaluate the 
   seismic adequacy of horizontal tanks consistent with engineering practice 
   and, therefore, acceptable for existing installations only.

Conclusion

On the basis of its review of Section 7 of Part II of GIP-2, the staff 
concludes that the methodology provided for the seismic adequacy evaluation 
of the safety-related horizontal and vertical tanks and heat exchangers 
existing at the USI A-46 plants is acceptable.  However, the criteria for 
evaluating tanks and heat exchangers, as defined herein, are not acceptable 
for new installations.

II.8  Cable and Conduit Raceway Review

Discussion

Section 8 of Part II of GIP-2 describes the screening guidelines for cable 
and conduit raceway review.  The screening procedure is based primarily on 
earthquake experience data and some shake-table test data.  Several types of 
raceway configurations and support systems are covered in this section.  The 
guidelines consist of a set of walkdown guidelines and a set of limited 
analytical review guidelines.

The walkdown guidelines provide guidance for the seismic review teams (SRTs) 
to: (1) perform direct in-plant screening reviews of raceway systems against 
a set of inclusion rules, (2) assess other seismic performance concerns not 
covered by the inclusion rules, and (3) select, during the walkdown, 10 to 
20 representative, worst-case samples of raceway supports for analytical 
review.  The systems which are identified to be within the boundaries of the 
inclusion rules would be considered to be within the applicability limits of 
the experience database.  If violations of the inclusion rules are observed, 
the SRT should investigate the specific conditions of the cable tray systems 
with proper assessment methodology to verify their seismic adequacy. 

.

The purpose of the limited analytical review is to ensure that the selected 
worst-case, representative samples of the raceway support systems in the 
plant are at least as rugged under the required seismic loadings as those in 
the earthquake experience and shake-table test databases that performed 
well.  Section 3.3 of GIP-2 Reference 9 should be used for selecting samples 
for the limited analytical review.  If these samples do not pass this 
limited analytical review, further evaluations should be conducted and the 
sample should be expanded as appropriate.  The analytical reviews are 
primarily based on the back-calculated capacities of raceway supports in the 
seismic experience database.  They are formulated with the use of static 
load coefficients, plastic behavior structural theory, and professional 
engineering judgment to ensure that cable tray and conduit supports are 
seismically adequate and as rugged as those in the seismic experience 
database.  The main feature of the reviews is that all supports in the 
selected worst-case samples are checked for deadload (DL) vertical capacity 
using the working stress criteria given in Part 1 of the American Institute 
of Steel Construction (AISC) Specification.  All supports in the selected 
worst-case samples must pass the DL check, otherwise the supports must be 
treated as outliers and disposed of as such.  However, isolated cases of a 
support not meeting the one DL criterion could be accepted if the raceway 
support system has high redundancy; this can be demonstrated by showing that 
the adjacent supports are capable of satisfying the walkdown guidelines, 
including the inclusion rules and the analytical review guidelines.  In 
addition to the DL check, all of the cable tray supports in the selected 
worst-case samples suspended from overhead must satisfy three times the DL, 
otherwise the supports must be treated as outliers.  This check is designed 
to ensure that the anchorage supporting the cable trays and conduit raceway 
in the USI A-46 plants is as strong as those in the experience database in 
sustaining the vertical loads.

The raceway hardware becomes an outlier if it does not meet the walkdown 
guidelines (inclusion rules and other seismic performance concerns), or the 
limited analytical review guidelines.  When an outlier is identified, 
additional evaluations as described in GIP-2 Reference 9, or alternative 
methods, are required to demonstrate seismic adequacy of the raceway 
hardware and to resolve the outlier issue.  The evaluations and 
justifications to be used to resolve the outlier issue should be based on 
mechanistic principles and sound engineering judgment and should be 
thoroughly documented for NRC staff review.

Evaluation and Conclusion

The staff has reviewed the guidelines proposed by the SQUG for evaluating 
the seismic adequacy of cable and conduit raceway systems.  The main 
objective of the proposed guidelines was to develop a cost-effective means 
of verifying the seismic adequacy of raceway supports in USI A-46 plants.  
These guidelines were developed on the bases of analytical studies, 
shake-table experimental model tests, and assessment of the performance of 
cable and conduit support systems in past earthquakes.

The staff considers that the plant walkdown guidelines represent an 
acceptable approach for evaluating the seismic adequacy of existing cable 
and conduit raceways in USI A-46 plants.  Also, the staff agrees that the 
proposed 
.

analytical procedure is a reasonable approach to ensure that the cable and 
conduit raceways and supports in USI A-46 plants, when all the guidelines 
are satisfied, are as rugged as those observed in the past earthquake 
experience data.  Although the proposed guidelines would not require 
detailed analyses and, therefore, would not predict the structural response 
of the raceway support systems, they should provide the needed rationale to 
judge the seismic adequacy of the raceway support systems with a reasonable 
factor of safety.  Therefore, the staff concludes that the proposed 
guidelines for evaluation of seismic adequacy of cable and conduit raceways 
and their supports are acceptable subject to the staff evaluations described 
in this supplement.

II.9  Documentation

Section 9 of Part II describes the documentation that is to be submitted to 
the staff upon completion of the plant-specific review and includes the 
documentation available at the plant site for audit.  The major document 
types are:

   safe-shutdown equipment list report 
   relay evaluation report 
   seismic evaluation report 
   completion letter

The staff has reviewed the outlines of each report as given in GIP-2.  The 
information to be submitted to NRC for review will provide overall results 
of the implementation program.  Therefore, the staff finds the proposed 
plant-specific information to be submitted to the NRC for resolution of USI 
A-46 acceptable.

However, GIP-2 recommends documentation (not required to be submitted to the 
NRC) of only the results from several evaluations (e.g., Sections 9.3 and 
9.4) and not the assumptions and judgments used for the respective 
evaluations.  The staff recommends documentation of the assumptions and the 
judgments as previously mentioned in Section II.2 of this supplement.  The 
documentation of assumptions and judgments, in addition to the results of 
evaluations, will facilitate the reconstruction of relevant basis for the 
licensee's evaluations.

II.10  References

Section 10 of Part II contains a list of references that are the source of 
information for the criteria and procedures described in GIP-2.  During the 
course of its review, the staff consulted References 5, 6, 7, 8, 9, 10, 26, 
32, and 33, among others, of GIP-2, in order to develop the bases for 
accepting the criteria and procedures presented in GIP-2 for implementing 
USI A-46 resolutions.  Therefore, the evaluations and conclusions presented 
in this SSER No. 2 are based on the information provided in each reference 
as dated in GIP-2 with the exception of Reference 4 for the reasons stated 
below.  If any updated references are to be used for the USI A-46 program, 
they must be submitted for staff review and approval.


.

As noted in Section II.5 of this supplement, Reference 4 of GIP-2 has been 
modified extensively by the addition of unreviewed new information.  Because 
the staff has not reviewed this particular version of the reference, any 
specific application of the detailed information documented in Reference 4 
for the implementation of USI A-46 should be submitted to the staff for 
review and approval before it is used.  For the evaluation of the 
acceptability of new data, see the staff position described in Section I.3.0 
of this supplement.

III  APPENDICES

III.1   Appendix A, Procedure for Identification of Safe-Shutdown Equipment

Appendix A of GIP-2 amplifies the method described in Section 3 of Part II 
for identifying safe-shutdown equipment.  The staff incorporated its 
evaluation of Appendix A into its discussion in Section II.3 of this 
supplement.

III.2   Appendix B, Summary of Equipment Class Descriptions and Caveats

Appendix B of GIP-2 incorporates information regarding the seismic 
capacities of 20 equipment classes.  This information was extracted 
principally from GIP-2 Reference 5 and partially from GIP-2 Reference 4 for 
earthquake experience data, and from GIP-2 Reference 6 for the test data.  
The staff evaluation of this appendix shall be used in conjunction with the 
staff evaluations presented in Sections II.4, II.5, and II.6 of this 
supplement.

In GIP-2 Reference 5, SSRAP documented its review of GIP-2 Reference 4, 
GIP-2 Reference 6, and other supporting documents.  After a detailed and 
careful review of the full range of the available experience database, 
combined with the general experience of the SSRAP members, the SSRAP 
concludes that the equipment (20 classes) presented in Appendix B of GIP-2, 
when properly anchored, and with some reservations as discussed in GIP-2 
Reference 5 and Appendix B of GIP-2, have an inherent seismic ruggedness and 
a demonstrated capability to withstand seismic motion bound as specified 
without significant structural damage and malfunction.  The staff concurs 
with this conclusion.

On the basis of the discussion described above and the review of information 
presented in Appendix B of GIP-2 and other supporting documents, the staff 
concludes that Appendix B is generally acceptable, subject to the following:

1. Throughout Appendix B, such statements as "equipment determined to be 
   seismically rugged" are repeatedly used.  The staff considers such 
   statements ambiguous unless the appropriate vibration level for which the 
   equipment is rugged is given.  In addition, the first sentence of each 
   equipment class states that the equipment "has been determined to be 
   seismically rugged...provided the intent of each of the caveats listed 
   below is met...."  The staff also finds such statements to be incomplete 
   and misleading because according to Appendix B, a user can simply meet 
   the caveats and declare the equipment to be rugged (and therefore 
   acceptable) for its application without even comparing it with the demand 
   vibration level.  The staff takes the position that in addition to 
   meeting the caveats, the user must demonstrate the demand level is 
   appropriately satisfied by the capacity level before the 
   .   

   equipment can be considered to be rugged and acceptable for its 
   application.

2. Regarding the attachment weight of 100 pounds, GIP-2 uses the term "a 
   cabinet assembly" (e.g., page B.1-4, MCC/BS caveat 4; page B.2-3, LVS/BS 
   caveat 5).  The staff understands this term to mean a combination or 
   lineup of a number of individual cabinets, bays, or frames.

3. GIP-2 includes a caveat for many equipment classes that the sections of 
   the multibay cabinet should be bolted together only "if any of these 
   cabinets contain essential relays...."  Since the database cabinets in 
   GIP-2 Reference 6 were bolted during testing, the adjacent cabinets at 
   the plants should be bolted for applicability of the GERS level, even 
   though these cabinets do not contain relays.  Otherwise, the responsible 
   review engineer should justify the use of GERS level for those cabinets.

4. The capacity levels for motor operators on valves presented in GIP-2, and 
   in GIP-2 Reference 6, appear to be high compared to the levels reported 
   in NUREG/CR-4659, Vol. 4 (Reference 10).  Note that new data from this 
   report were not available to the SQUG at the time GIP-2 was being 
   prepared.  

   Therefore, the information presented in Reference 10 should be considered 
   in the evaluation of motor operators on valves, especially for some 
   earlier models.

III.3  Appendix C, Anchorage Data

Appendix C of GIP-2 includes information necessary for verifying the 
adequacy of anchorage.  The staff incorporated its evaluation of this 
appendix in Section II.4 of this supplement.

III.4  Appendix D, Seismic Interaction

Appendix D of GIP-2 describes seismic interaction as the physical 
interaction of any structures, piping, or equipment with nearby 
safe-shutdown equipment caused by relative seismic motions.  Three seismic 
interaction effects are covered in the GIP, namely:  proximity, structural 
failure and falling, and flexibility of the attached lines and cables.  The 
staff finds the guidelines for reviewing seismic interaction adequate and, 
therefore, this appendix is acceptable and should be used with the staff 
evaluation presented in Section II.4 of this supplement.

III.5  Appendix E, Preparatory Work Prior to Walkdown

Appendix E of GIP-2 describes the experience gained from previous walkdowns 
to maximize the effectiveness of the walkdown.

The GIP states that most of the equipment "has been shown to be seismically 
rugged...."  As explained in Section III.2 above, the staff considers this 
statement ambiguous unless the appropriate vibration level is associated 
with it.
.

III.6  Appendix F, Screening Walkdown Plan

Appendix F of GIP-2 describes the organization and approach that can be used 
by the seismic review team, the degree of inspection to be performed, the 
walkdown logistics to be followed, and the screening walkdown to be 
completed.  The staff finds the screening walkdown plan adequate for 
accomplishing the objectives of the walkdown inspection and, therefore, the 
staff concludes that the GIP screening walkdown plan is acceptable.

III.7  Appendix G, Screening Evaluation Work Sheets

Appendix G of GIP-2 provides the work sheets to be used by the walkdown team 
to document their review.  The staff finds these work sheets to be a useful 
summarized checklist that briefly documents the responses to essential 
screening questions addressing the seismic adequacy of each piece of 
equipment and, therefore, the staff concludes that this appendix is 
acceptable subject to the following staff positions:

1. Since the screening evaluation work sheets in Appendix G contain summary 
   information presented in Appendices B and C, in case there is any 
   conflict between these pieces of information, the information in 
   Appendices B and C should be used.  For example, for motor control 
   centers, the weight of 800 pounds should be considered maximum instead of 
   average (pages B.1-7 and G.1-2).

2. The screening evaluation work sheets do not require documentation of 
   manufacturer, model, etc.  For information purposes only, the staff 
   strongly recommends that such information should be recorded if readily 
   available.

3. Since GIP-2 allows the demand level to exceed the capacity level under 
   certain conditions, in response to the question "Does capacity exceed 
   demand?" on the Screening Evaluation Work Sheet (SEWS) for each piece of 
   equipment, the reviewer must also identify whether the exceptions 
   described on page 4-10 of GIP-2 were used in the comparison (see Item 1 
   of Section II.4.2 of this supplement).  


CONCLUSION

The staff concludes that GIP-2, dated February 1992 (Reference 8), 
supplemented by the staff positions, clarifications, and interpretations, 
stated herein for each section of GIP-2, constitutes an acceptable method 
for the implementation of the resolution of USI A-46 as specified in Generic 
Letter 87-02, subject to the following exceptions: 

1. Section II.3, Identification of Safe-Shutdown Equipment: Evaluation and 
   Conclusion, Item 4

   The NSSS equipment exclusion given in Section 3.3.2 of GIP-2 does not 
   apply to safety-relief valves included in the scope of USI A-46.

.

2. Section II.4.1, SQUG Commitments

   If the licensee commits to use GIP-2 for the implementation of USI A-46, 
   it must commit to both the SQUG commitments and the use of the entire 
   implementation guidance provided in GIP-2, unless otherwise justified to 
   the staff as described in GIP-2 and this supplement.

3. Section II.4.4, Anchorage Adequacy, Item 9, Verification of Anchorage 
   Capacity by Computer Codes

   The EBAC code as given in Reference 7 of GIP-2 and the ANCHOR code as 
   given in Reference 14 of GIP-2 are acceptable provided that the items of 
   concern discussed in Item 9 of Section II.4.4 of this SSER No. 2 are 
   adequately considered prior to their applications.

4. Section II.4.4, Anchorage Adequacy, Item 10, Verification of Anchorages 
   in Inaccessible Areas

   To ensure relay functionality, the licensee should try all practicable 
   means to inspect all of the anchorages of cabinets having essential 
   relays.

5. Section II.4.4, Anchorage Adequacy, Item 11, Minimum Spacing Between 
   Anchors

   On page 4-39 of GIP-2, the sentence, "The minimum spacings given in 
   Appendix C are for distances between adjacent anchors in which the cones 
   of influence just touch each other at the surface of the concrete..." is 
   incorrect.  The minimum spacing values in GIP-2, Appendix C correspond to 
   a 13-percent shear cone overlapping.  The quoted statement should be 
   corrected.

6. Section II.4.4, Anchorage Adequacy, Item 12, Use of ACI 349

   The use of ACI 349, Appendix B, for the resolution of anchorage outliers 
   is not acceptable to the staff.

7. Section II.5, Outlier Identification and Resolution, and Section II.10, 
   References

   The staff has not reviewed the current version of GIP-2 Reference 4.  Any 
   specific application of this reference for resolving USI A-46 should be 
   submitted to the NRC staff for review and approval before it is used.

8. Section II.6, Relay Functionality Review, Item 1, Use of Zero Period 
   Acceleration Criteria

   For the relay functionality review, in addition to the comparison of the 
   spectral accelerations, the ZPA capacities should be compared and shown 
   to be adequate.


.

9. Section III.2, Appendix B, Summary of Equipment Class Descriptions and 
   Caveats, Item 3

   GIP-2 includes a caveat for many equipment classes that states that the 
   adjacent sections of multibay cabinets should be bolted together only "if 
   any of these cabinets contain essential relays."  For multibay cabinets, 
   all cabinets, regardless of whether or not a cabinet contains essential 
   relays, should be bolted together in order for the GERS to be applicable, 
   unless justification can be provided for applying these GERS to multibay 
   cabinets that are not bolted together.

10.   Section III.2, Appendix B, Summary of Equipment Class Descriptions and 
   Caveats, Item 4

   The information presented in Reference 10 of this SSER No. 2 should be 
   considered in the evaluation of motor operators on valves, especially for 
   some earlier models.

11.   Section III.7, Appendix G, Screening Evaluation Work sheets, Item 1

   In any case where the information presented in Appendix G conflicts with 
   the information in Appendices B and C, the information in Appendices B 
   and C should be used.

12.   Section III.7, Appendix G, Screening Evaluation Work sheets, Item 3

   In response to the SEWS question, "Does capacity exceed demand?", the 
   reviewer must identify whether the exceptions described on page 4-10 of 
   GIP-2 were used in the comparison.
.

REFERENCES

1.    Generic Letter 87-02, "Verification of Seismic Adequacy of Mechanical 
      and Electrical Equipment in Operating Reactors, Unresolved Safety 
      Issue (USI) A-46," U.S. Nuclear Regulatory Commission, February 19, 
      1987.

2.    "Generic Implementation Procedure (GIP) for Seismic Verification of 
      Nuclear Plant Equipment," Revision 0, Seismic Qualification Utility 
      Group, June 1988.

3.    "NRC Generic Safety Evaluation Report on the Seismic Qualification 
      Utility Group Generic Implementation Procedure, Revision 0, for 
      Implementation of USI A-46," July 29, 1988.

4.    "Generic Implementation Procedure (GIP) for Seismic Verification of 
      Nuclear Plant Equipment," Revision 1, Seismic Qualification Utility 
      Group, December 1988.

5.    "Supplemental Safety Evaluation Report No. 1 on SQUG Generic 
      Implementation Procedure, Revision 1," by NRC, June 29, 1990.

6.    "Generic Implementation Procedure (GIP) for Seismic Verification of 
      Nuclear Power Plant Equipment," Revision 2, Seismic Qualification 
      Utility Group, September 1990.

7.    Letter, James A. Norberg, NRC, to Neil P. Smith, dated March 11, 1991, 
      "Comments on Revision 2 of Generic Implementation Procedure (GIP) for 
      Use in USI A-46 Programs, dated September 21, 1990."

8.    "Generic Implementation Procedure (GIP) for Seismic Verification of 
      Nuclear Power Plant Equipment," Revision 2, Corrected 2/14/92, Seismic 
      Qualification Utility Group, February 1992.

9.    Letter, W. A. Von Reisemann, SSRAP, to Pei-Ying Chen, NRC, dated July 
      23, 1991, "EQE Class of Twenty Report."

10.   NUREG/CR-4659, Vol. 4, "Seismic Fragility of Nuclear Power Plant 
      Components (Phase II):  A Fragility Handbook on Eighteen Components," 
      June 1991.
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