ML20249C097
| ML20249C097 | |
| Person / Time | |
|---|---|
| Issue date: | 06/19/1998 |
| From: | Dan Dorman NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR NUDOCS 9806260002 | |
| Download: ML20249C097 (92) | |
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UNITED STATES B
NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20086 4 001
\\,w +/
June 19, 1998
SUBJECT:
SUMMARY
OF MEETING BETWEEN NRC STAFF AND SEISMIC QUALIFICATION UTILITY GROUP Pursuant to notice, the NRC staff met with representatives of the Seismic Qualification Utility Group (SQUG) on May 18,1998, at the NRC headquarters office in Rockville, Maryland. The meeting was requested by SQUG to discuss three issues which SQUG considers to be significant in relation to the implementation of the resolution of unresolved safety issue (USI)
A-46 at affected plants and to the future use of the Generic Implementation Procedure, I
l Revision 2 (GIP-2), for determining the seismic adequacy of new and replacement equipment (NARE) at these plants. These issues are not new; however, they represent the remainder of a iarger population of issues that has been the subject of many interactions and correspondence
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betwNa the staff and SQUG in the last two years.
A list of persons attending the meeting is provided in Enclosure 1. Enclosures 2 through 5 contain information presented by SQUG at the meeting.
Seismic Demand The first issue is reisted to the differing positions between SQUG and the staff on the proper selection of seismic demand for evaluating the seismic adequacy of equipment located inside structures at elevations within 40 feet above grade level. SQUG asserts that GIP-2 allows the use of ground response spectrum to determine the seismic demand for any equipment located within 40 feet above grade levelin typical stiff nuclear plant structures. SQUG believes that inherent conservatism in the calculational methods employed in developing in-structure spectra leads to unrealistic high amplifications of input ground spectra. The staff disagrees with SQUG's assertion on two fundamental points. The text in GIP-2 (page 4-16) clearly imposes a restriction and limitation on the use of ground response spectrum for seismic demand to conditions where the amplification factor between the free-field response spectra and in-structure response 1
spectra will not be more than about 1.5. Even in the absence of such clear restriction in GIP-2, the staff finci, it technically unjustifiable to use ground response spectra to determine the seismic demand (within 40 feet above grade)in a structure for which the licensee has developed specific in-structure response spectra wellin excess of 1.5 times the ground spectrum. In many cases, l
the spectra in question is a revised spectra that was proposed by the licensee, and accepted by the staff, for the purpose of the USl A-46 resolution, that is less conservative than the licensing basis spectra as a result of using the more refined analytical assumptions and techniques permitted by the current Sr.?. Discussion of this issue was concluded with commitment from SQUG to develop a stronger argument in support of the appropriateness of its interpretation of j
the GIP-2 provision for determining equipment seismic demand. [SQUG's response was i
provided in a letter from Neil P. Str.ith (SQUG) to Brian W. Sheron (NRC) dated June 10,1998.
This letter was the subject of a June 19,1998, teleconference between the NRC and SQUG.
During the teleconference, SQUG agreed to supplement the June 10,1998, letter with certain plant-specific information regarding licensee implementation of this aspect of GIP-2.)
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Ii Cable Tray Dudihty i
l SQUG had expressed concem regarding NRC requests for additional loformation (RAls) sent to
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some licensees regarding ductilitty of cable and raceway supports SQUG viewed the staff requests for lists of all raceways determined to be ductile during plant walkdowns and for calculated maximum ductility of the weakest raceway supports to be beyond the scope of GIP-2, as appmved, and a change of staff position. The staff agreed to clarify the approach in their requests as part of the meeting summary.
GIP-2 excludes horizontal response evaluations for raceway and cable tray supports which are judged to be ductile. The NRC staff reviewed and accepted the raceway and cable tray support verification approach provided in GlP-2. The staff concem is not necessarily with the GlP-2 concept but rather with the way it has been employed by some licensees. Some raceway and cable tray configurations in USl A-46 plants are not in the SQUG experience database and may have been inappropriately declared to be ductile. In some cases where cable trays have rigid lateral supports that are not ductile, licensees have stated that the non-ductile cable tray supports would eventually become ductile by inelastic deformation, buckling or failure of the rigid lateral members. Because of concems about these conditions, licensees are being asked to explain and justify their determinations, mado during plant walkdowns, that the raceway and cable tray supports are ductile. They are also being asked what percentage of their raceway and cable tray supports have been declared to be ductile. The SQUG states that GlP-2 and training contain the criteria and guidelines for determining whether raceways and cable trays are covered by the earthquake experience data and whether supports are ductile or not. Although the staff accepts the SQUG position that raceway and cable tray supports in the database are inherently rugged, for A-46 plants with raceway or cable tray configurations which are not in the experience database or those with obviously rigid support configurations, the staff will continue to ask licensees to justify certain adequacy determinations.
Questions about raceway and cable tray supports asked earlier of a few licensees were broad in scops and were intended to elicit information needed for the staff to understand the logic used by individuallicensees to evaluate supports that are'not ductile. The staff has also asked for ductility assessments in order to determine the reasonableness of the licensee's disposition of non-ductile or out-of-database supports. The staff, as a result of interactions with SQUG, has simplified its questions to focus only on the implementation of GIP-2 for cable tray and raceway supports at individual plants. This means that: a) questions asking for horizontal response
)
analyses are eliminated, b) maximum ductility evaluations are not needed, and c) listing all ductile raceway and cable tray supports is not required. Information related to the percentage of supports dispositioned by'a ductility conclusion may still be requested in some situations.
New and Replacement Enuinment/ Incorporation of GlP-2 into Farmtv Licensing Bases The finalissue of concem to SQUG is its perception of apparent change in the staff position on
' future applications of GlP-2 by affected utilities for NARE and the means by which it could be incorporated into the licensing basis. SQUG believes that GlP-2 can be incorporated into the licensing basis for a facility via a simple 50.5g evaluation, on the premise that the application of the GlP-2 methodology for the USl A-46 program has resulted in an overall improvement to plant b
l 1
l safety. SQUG believes that incorporation of GIP-2 by 50.59 should not trigger an unreviewed safety question (USQ) if the methodology is viewed as a whole approach and not broken down into various components. On the subject of timing of the licensing basis revision, the staff reminded SQUG of the conditiori stated in the staff's SSER No. 2 on GIP-2 in 1992, that consideration of GIP-2 incorporation into the licensing basis should be made after issuance of the final plant-specific safety evaluation resolving USl A-46. In regard to the means by which a
-licensing basis revision could be made, the staff committed to revisit the approach proposed by i)
SQUG for performing the 50.59 evaluations on a programmatic level rather than evaluating the various aspects of a specific program against existing licensee commitments in the FSAR. The staff's conclusions on this issue will be provided to SQUG in separate docketed correspondence.
original signed by Daniel H. Dorman, Lead Project Manager Division of Reactor Projects - 1/Il Office of Nuclear Reactor Regulation
Enclosures:
- 1. List of Attendees
- 2. Overview and Status of SQUG Program for Resolution of USI A-46 and Maintenance of Seismic Adequacy
- 3. References Associated with issue #1, Use of GlP Method A
- 4. References Associated with issue #2, Justification for Use of Cable Tray Methodology
- 5. References Associated with issue #3, Adoption of GlP as a Licensing Basis Method cc w/encls: See next page DISTRIBUTION M CoDV Central File PUBLIC PDI-2 Reading DDorman
- ACRS E-Mail SCollins/FMiraglia MO'Brien
. KManoly' JMoore, OGC BBoger '.
GBagchi RRothman '
TMartin (E-mail SLM3)
JZwolinski RWessman "
JStrosnider RCapra Glaines BSheron OFFICE DRPE/ LPM PDi-24,An EMEB/Bck PDI-2/D s;
NAMEi DDorman:stN MO'MI RkN RCapra V DATE ~
k/N/98 b /k/98 '
Y/98 6 //7/98
. OFFICIAL RECORD COPYi L
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4
_ _. _ _ _. safety. SQUG believes that incorporation of GlP-2 by 50.59 should not trigger an unreviewed safety question (USQ) if the methodology is viewed as a whole approach and not broken down into various components. On the subject of timing of the licensing basis revision, the staff reminded SQUG of the condition stated in the staffs SSER No. 2 on GlP-2 in 1992, that consideration of GlP-2 incorporation into the licensing basis should be made aBar issuance of the final plant-spoofic safety evaluation resolving USl A-48. In regard to the means by which a
'icensing basis revision could be made, the staff committed to revisit the approach proposed by SQUG for performing the 50.59 evaluations on a programmatic level rather than evaluating the various aspects of a speci'ic program against existing licensee commitments in the FSAR. The l
staffs conclusions on this issue will be provided to SQUG in separate docketed correspondence.
j CD Daniel H. Dorman, Lead Project Manager Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Enclosures:
1 List of Attendees
- 2. Overview and Status of SQUG Program i
for Resolution of USl A-46 and I
Maintenance of Seismic Adequacy
- 3. References Associated with issue #1, Use of GlP Method A
- 4. References Associated with issue #2, Justification for Use of Cable Tray Methodology
- 5. References Associated with issue #3, Adoption of GIP as a Licensing Basis Method cc w/encis: See next page
cc list:
Mr. Neil P. Smhh, Chairman.
Seismic Qualification Utility Group do MPR Associates, Inc.
320 King Street Alexandria,VA 22314 Mr. R. Kassawara EPRI Program Manager 3412 Hillview Avenue P.O. Box 10412 Palo Alto, CA 94304 I
i l
l.
l
LIST OF ATTENDEES MEETING BETWFEN NRC AND THE SEISMIC QUALIFICATION UTILITY GROUP MAY 18.1988 1
NAME ORGANIZATION D. Dorman NRC/NRR R. Wessman NRC/NRR G. Lainas NRC/NRR K. Manoly NRC/NRR R. Rothman NRC/NRR G. Bagchi NRC/NRR J. Strosnider NRC/NRR B. Sheron '
NRC/NRR J. Moore NRC/OGC N. Smith SQUG M. Burzynski SQUG J. Fisicaro SQUG W. Schmidt SQUG D. McCombs SQUG E. Forrest NUSIS A. Wyche Bechtel i
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Copies of References Associated With issue #1 Use of GIP Method A (Documents Arranged in Order of Reference Number) i E_-__-__---_--_---------------._
IR IOSURE 3
GL 82-0 2 Enclos vra,2 llAl27
~
y Lf y Equipment Class Bound Motor-operated valves with large eccentric-operator-Type C 1engths-to pipe-diameter ratios y
Motor-operated valves (exclusive of those with large eccentric-operator-J lengths-to pipe-diameter ratios)
Air-operated valves Type A Horizontal pumps and their motors Vertical pumps and their motors n-
~
cThese spectrum bounds are intended for comparison with the 5% damped design
\\,7
, horizontal around response spectrum at a given nuclear power plant.
In other words, if the horizontal ground response spectrum for the nuclear plant site is less than a bounding spectrum at the approximate frequency of vibration of the equipment and at all greater frequencies (also referred to as the frequency f
range of interest), then the equipment class associated with that spectrum is considered to be included within thn scope of this method. Alternately, one S
may compare 1.5 times these spectra with a given 5% damped horizontal flo.or N
[spectruminthenuclearplant.
i The comparison of these seismic bounds with the design horizontal ground response spectrum is judged to be acceptable for equipment mounted less than about 40 feet *
)
q(istiffstructures.above grade (the top of the ground surround For equipment mounted more than about 40 feet above grade, y
/
qcomparisonsof1.5timesthesespectrawiththehorizontalfloorspectrumis In all cases such a comparison with floor spectra is also acceptable.
necessary.
J The vertical component will not be any more significant relative to the horizon-There-
/talcomponentsfornuclearplantsthanitwasforthedatabaseplants.
I fore, it was decided that seismic bounds could be defined purely in terms of
/
f horizontal motion levels.
)
~
The criteria are met so long as the 5% damped horizontal design spectrum lies
/ below the appropriate bounding spectrum at frequencies greater than or equal
/
/tothefundamentalfrequencyrangeoftheequipment.p judgment 311y This estimate can be made The recommendation that the seismic bounding spectrum cen be compared with the horizontal design ground response spectrum for equipment mounted less than abouti 40 feet above grade is based upon various judgments concerning how structures respond in earthquakes. However, this 40-foot above grade criterion must be applied with some judgment because some structures may respond in a different manner.
(2) Motor Control Centers Motor control centers contain motor starters (contactors) and disconnect switches. They als) provide over-current relays to protect the system from In most cases where numerical values are given in this section they should be considered as either " approximate" or "about," and a tolerance about the stated value is implied.
9 Enclosure
r
?
S'SRAP Sqo/f February 28, 1991 two-thirds of the estimated average free-field ground motion to which the data base equipment was actually exposed at sites with estimated mean peak ground accelerations in excess of about 0.4g.
The derivation of this Bounding Spectrum is discussed in more detail in Appendix A.
The generic t
Bounding Spectrum is defined in terms of the 5% damped horizontal ground I
j response spectrum in Figure 3.1 and Table 3.1.
This Bounding Spectrum must be used together with the caveats and exclusions listed in the subsequent sections for each of the classes of equipment.
In the case of some medium voltage switchgear and some motor-operated valves with large E
{
eccentric operator length to pipe diameter ratios, this Bounding Spectrum E
l may be factored downward as described in Sections 10C and 20 respectively.
This spectrum bound is intended for comparison with the 5% damped f
i design horizontal cround response spectrum at a given nuclear power plant.
I In other words, the experience data base demonstrates adequate seismic l
1 I
ruggedness when the horizontal ground response spectrum for the nuclear i
plant site is less than the Bounding Spectrum at a reasonably conservative i
j I
lower bound estimate of the fundamental frequency of vibration of the i
equipment and at all greater frequencies (also referred to as the frequency range of interest).
Alternately, one may compare 1.5 times the Bounding Spectrum with a given 5% damped horizontal in-structure spectrum in the nuclear plant over this frequency range of interest.
The comparison of the seismic bound with design horizontal ground 1
response spectra is judged to be acceptable for equipment with frequencies
- In most cases where numt rical values are given in this report, they should be considered as either " approximate" or "about" and a tolerance about the stated value is implied.
1-15
hlb-L Revision 2 Corrected,6/28/91 4.1.6 Documentation The licensee will document the results of the Screening Verification and Walkdown on Screening Verification Data Sheets (SVDS).
4.2 SEISMIC CAPACITY COMPARED TO SEISMIC DEMAND The first screening guideline which should be satisfied to verify'the seismic adequacy of an ' item of mechanical 'or electrical equipment.is to confirm that the seismic capacity of the equipment is greater than or equal to the seismic demand imposed on it.
This section addresses the comparison of seismic capacity to seismic demand for the eouioment itself. Note that a comparison of seismic capacity to seismic demand is also made for the anchorace of the equipment in Section 4.4 and for the relays mounted in the equipment in Section 6.
7
'% )
The seismic capacity of equipment can be represented by a " Bounding Spectrum" based on earthquake experience data, or a " Generic Equipment Ruggedness Spectrum" (GERS) based on generic seismic test data. These two types of seismic capacity spectra are described in Sections 4.2.1 and 4.2.2, respectively. Note that these two methods of representing seismic capacity of equipment can only be used if the equipment meets $he intent of the caveats for'its equipment class as described in Section 4.3.
eismic capacity of an item of equipment can be compared to a seismic
. demand which is defined in terms of either a ground response spectrum or an i in-structure response spectrum. Table 4-1 outlines these types of hcomparisonsasefferMethodAorB. Method A is for making a comparison hthagroundresponsespectrum;Section4.2.3 y
4-8
Refff S S fi A F R,p,7-February 28, 1991 obtain amplifications greater than about 1.5 at frequencies above about 8 Hz even at low elevations in stiff structures.
However. SSRAP notes that high frequency ground motions documented to date do not have much damage Potential due to low spectral displacement, low energy content and short t
duration.
Further, the equipment of concern does not appear to have a significant sensitivity to high frequencies (other than possibly relay chatter).
Thus, the issue of high frequency amplitudes through buildings I
is only a concern for the potential failure of brittle materials such as
/
ceramics and for potential chatter.
It is SSRAP's firm opinion that the issue of potential amplifications greater than 1.5 above about 8 Hz for high frequency input is of no consequence for the classes of equipment
\\
considered in this document except possibly for relay chatter.
The SSRAP recommendations for the resolution to the questions of the vulnerability of relays to chatter due to higher frequency input and the amplification J
of high frequency input are discussed in Section 5.
For functionality l
/
after the earthquake and structural integrity of these_ classes of S
equipment, SSRAP considers its ruggedness recommendations to be
/ particularly conservative for high frequency earthquake inpucs so long as y
(,
the Bounding Spectrum is not exceeded.
~_:-_-
A.3 VERTICAL COMPONENT OF MOTION SSRAP believes that the vertical component will not be more si nificant relative to the horizontal components for nuclear plants than 5
it was for the data base plants.
Therefore, it was decided that seismic i
bounds could be defined purely in terms of horizontal motion levels.
c.
j
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I-106
1 i
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i February 28, 1991
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insulation, this elevation would correspond to the bottom of the crushable
?
foam insulation.
For the case of internal structures not laterally tied r
to external valls, grade should be considered to be the base of the (l
internal structure.
For softer soil sites where one does expect substantial deamplification with depth, the top of the ground surface should be considered to represent grade, even for tho above two i
exceptions.
In other words, judgment must be exercised when interpreting f
9 the "effectiveN grade elevation, and in some cases this elevation should
[.
be considered to be lower than the top of the ground surrounding the s
(
building.
}
j Similarly, unique conditions could exist where amplification greater \\
(:
l y
than about 1.5 would be expected to occur within 40 feet ah ve grade.
t SSRAP is unable to come up with any realistic example from nuclear power i
plant structures where greater amplifications would be expected.
Even so, the provision of allowing comparison of the seismic bounds to the design horizontalgroundresponsespectraforequipmentmountedlessthan40feg
~
above grade should be applied with judgment.
e -_
/
The above discussion does assume that the ground motion will have frequency content similar to that represented by the four data base j spectra and the Reference Spectrum in Figure' A.1, or by either the U.S.
j (NRCReg. Guide 1.60 Spectrum (Reference 23) or the NUREG/CR-0098 Spectrum Reference 24).
For all of those records, 90% of the power comes from frequencies below about 7 Hz (see Reference 34).
Many lower magnitude i
I (less than magnitude 6.0) records associated with short epicentral ranges show significant power coming from frequencies in excess of 8 Hz, particularly those recorded in the eastern U.S.
In such cases, one might
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E 1-105 L
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'F i
28f,ff STEAP Recut February 28. 1991 ground motion.
For embedded structures on rock sites, one would not expect much deamplification of horizontal ground motion with depth.
On the other hand, if the structure is laterally supported by the surrounding rock, one would expect this surrounding rock to prevent amplification of motion within the structure.
Thus, one would expect horizontal motion within the structure' at and below grade to nearly correspond to the free-field horizontal ground motions because of this lateral support provided by the surrounding rock.
Next, with moderately stiff structures, as found in nucaear power plants, one would not expect amplification of motions by a factor greater than about 1.5 within a 40 feet elevation change.
- Thus, amplification of the horizontal free-field ground spectra by factors greater than 1.5 are considered to be generally unlikely for elevations less than 40 feet above grade.
There are exceptions to these expectations.
Some examples are:
1.
Sometimes the exterior side walls of an embedded structure are h
surrounded by crushable foam insulation so as to isolate the structure from lateral support by the surrounding rock or soil.
2.
Often the internal structure within the containment building is laterally tied to the external containment wall so that the not internal structure is not laterally supported by the surrounding rock or soil.
For rock or very stiff soil sites where one does not expect one should substantial deamplification of horizontal motion with depth, consider grade for each structure to be the highest elevation at which that structure is laterally supportod by the surrounding rock or soil.
For the case of external side walls surroundad by crushable foam I-104 a
C.[h-L Revision 2 Corrected, 6/28/91 from the surrounding structure, then the " effective grade" is the elevation at the top of the base mat.
l l
Advantaae and Limitations. The advantage of using ground response comparisons is that with the applicable restrictions and limitations, all the equipment covered by the Bounding Spectrum or the GERS can be evaluated )
l for seismic adequacy without the need for using in-structure response i
spectra which are often based on very conservative modeling techniques or
/
(
may not be available.
l The restrictions and limitations on use of the around response spectrum for comparisontotheBoundingSpectrumandtheGERSisbasedonthecondit I
that the amplification factor between the free-field response spectra and the in-structure resoonse spectra will not be more than about 1.5,'and that ;
the natural frequency of the equipment is not in the high energy range as I
follows:
t
/
n
\\j 1
The equipment should be mounted in the nuclear plant at an elevation
,l below about 40 feet above the effective grade, and i
The equipment, including its supports, should have a fundamental natural frequency greater than about 8 Hz.
i Seismic Capability r gineers should be alert for unusual, plant-specific n
situations which could cause the amplification factor to be greater than that of typical nuclear plant structures. The1.5amplifiEationfactoris only applicable to reinforced concrete frame and shear wall structures and i
to heavily-braced steel frame structures.
Guidance for Evaluatino In-line Eauioment. When using the Boundina Soectrum for the seismic capacity of equipment mounted on a piping system (i.e., valves, valve operators, and sensors), the 8 Hz limitation does ng1 apply.
That is, the piping system can have a natural frequency lower than about 8 Hz at the location where the item of equipment is mounted.
Low 4-16
l
$ 96[2 MO. 2.-
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4.
With regard to Section 3.3.2 of Part 11 ' 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 ccope because Reference 17 of GIP-2 does not provide a basis for excluding the safety-relief valves from the USI A-46 scope. '
Section 3.3.3 of Part II requires that any equipme'nt needed for safe 5.
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.
11.4 Screenino 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 D
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 i
visual inspection of the screening procedures. The evaluation of screening verification and walkdown follows in Sections 11.4.2, 11.4.3, and 11.4.4 of this supplement.
w, r ~lusion Conc l The staff has reviewed the screening procedures and criteria.
Based on the f evaluations and findings described in Sections !!.4.2,11.4.3, and 11.4.4 j
i 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, keeptlonsandpositionsdescribedinthesectionsthatfollow.
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.
t 12
S$ 62 Nu, 2.
- 'Y *0 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.
With respect to the " Definition of Terms #In the.last paragraph of 3.
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 l
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 l
response spectra (Category 1 plants without double asterisks Table A)
)
l l
and plants included in the Systematic Evaluation Program (SEP, Category s
l 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 f
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 l
f submit its schedule for implementing the resolution of USI A-46 within
\\
120 days after this supplement is issued. The plant-specific l
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 g plants, however, the 3-year period will not commence until one of the i
/ following conditions is met:
s (1) the receipt of staff approval of the in-structure response spectra to be used to resolve the USI A-46.
i I
( (2) 60 days following the licensee's initial submittal of acceptable procedures and criteria in generating those in-structure response 14 i
(
l NILC l< br M.ed g}b(q s a:
UNITED STATES g g/
c, y
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON. D.C. 'mann ung m %,
l f August 6, 1996 Mr. Neil P. Smith, Chairman Seismic Qualification Utility Group c/o MPR Associates, Inc.
320 King Street Alexandria, VA 22314
SUBJECT:
EVALUATION OF REVISION 3 TO THE GENERIC IMPLEMENTATION PROCEDURE FOR SEISMIC VERIFICATION OF NUCLEAR POWER PLANT EQUIPMENT (TAC NO. M93624)
Dear Mr. Smith:
By letter dated July 31, 1995, the Seismic Qualification Utility Group (SQUG) submitted the Generic Implementation Procedure, Revision 3 (GIP-3), for seismic verification of nuclear power plant equipment.
The NRC staff has completed its review of GIP-3 and its conclusions are documented in the enclosed safety evaluation (SE).
n The staff finds that the changes in the restrictions on the use of the U
in-cabinet amplification factor, corrections in equations for determining allowable anchorage capacity for horizontal tanks and heat exchangers, and the majority of the editorial and typographical changes are acceptable. However, the staff also finds that the changes in anchorage criteria and some editorial changes are not acceptable as delineated in the enclosed SE.
In consideration of revising the GIP, the staff has identified an area of concern regarding the criteria provided in GIP, Revision 2 (GIP-2), dated February I4, 1992.
The issue in question is that Section 4.2 of GIP-2 provides alternative criteria for the comparison of seismic demand with the seismic capacity, which may lead to a licensee's use of a demand spectrum less uservative than that specified in the_ plant's licensi_no basis documents.
e staff contends that the use of Method A in Section 4.2 of GIP-2 is only ]
appropriate for facilities that do not have in-structure response spectra in I
u their respective licensing basis documents.
The staff believes that the lack l of specificity in GIP-2 with regard to the selection of the appropriate method]
for determining equipment seismic adecuacy may_ lead to potential f
)
bonconformance.fThii! staff requests that SQUG respond to the staff's concerns and provide appropriate guidelines for the future revision to the GIP.
l O
N N. Lt Ytr~
ttk NIN b dotMFt Ed, y The staff does not consider a generic response to this question acceptable, and affected licensees should address this question on a plant-specific basis.
SQUG's Response in June 30.1997 Letter A comparison of the design ground response spectrum (GRS) to the SQUG bounding spectrum as a method for evaluating the seismic adequacy of equipment is included as Method A in Table 4-1 of GIP-2. The GIP allows this method to be used under the restriction that the equipment must be located at an elevation below about 40-feet above the effective grade of the building, gud the equipment fundamenta! frequency must be
' above about 8 Hz.
One of the advantages of the GlP Method A is that the various effects associated with in-structure responses are inherently included in the method. The GlP approach differs from current seismic licensing criteria in that, in GIP Method A, the seismic capacity of i
equipment and the seismic demand on this equipment are anchored to ground response spectra. Further, the GIP method is not based on the performance of a single item of i
equipment subjected to an artificial time history on a shake table. Instead the GlP method is based on the performance of numerous items of equipment subjected to several real earthquakes.
For these reasons, the GIP method, based on comparing ground response spectra at data base sites to safe shutdown earthquake (SSE) ground motion response spectra at l
nuclear plants, was accepted by recognized independent experts,i.e., Senior Seismic Review and Advisory Panel (SSRAP repon), including NRC staff, e.g., supplemental safety evaluation report (SSER) #2, as an acceptable method to verify the seismic adequacy of equipment installed in operating nuclear power plants.
In addition to the above, SQUG also provided some recommended screening method to SQUG members to address the issue.
SQUG's Response in November 11.1997 Letter SQUG's position is that the Method A option continues to be available (provided special conditions are met, i.e., the equipment is mounted below about 40 feet above the effective grade and its lowest natural frequency is above 8Hz) based on the clear rules in the GIP and it is applied property as clarified in SQUG's letter of June 30,1997, to the NRC on this subject. Any attempt to require use of 9nLy the 120-day spectra would rl represent a change in the NRC's published position and would have to be justified under l
NRC Staff Evaluation
l.
The SQUG-recommended three-part screening method discussed in the SQUG's June 30,1997, letter, does not address a major aspect of the staffs concem on this issue. In its discussion of GlP Method A contained in the June 30, and November 11, 1997, letters, SQUG failed to mention a limitation on the use of Method A contained in the GlP which is an essential consideration for the appropriateness of this method. On page 4-16 of GIP, Revision 2, in the Section entitled Advantaoe and Limitations. the GlP states
'The restrictions and limitations on use of the grgy.n.g response spectrum for comparison
N R C. Le b N d st/wlqn l
b t014r&
W.Q
. 3 --
to the Bounding Spectrum and the GERS is based on the conditions that the amplification factor between the free-field response spectra and the in-stnacture response spectra will not be more than about 1.5 and that the natural frecuency of the equipment is not in the hiah enerov range _..."[IEnrefore, me staff position is that if the floor response T spectrum in a structure is 1.5 or more times far or than_the free-field __ response spectrum,)
the. se of GIP Method A is not appropriate.
e RC staff will continue to ask each una u
A-46 plant licensee, where appropriate, the question contained in the staff's December 5, 1996 letter.
Furthermore, with regard to the issue of using the in-structure response spectra provided
' by licensees in their 120-day submittals, SQUG asserts, in the November 11,1997, letter, that requiring licensees to only use_those spectra _would have to be justified under 10 CFR 50.j109. The staff considers that these submittals constituted commitments from r ffectelYcensees regarding the quantification of the proposed seismic demand, to be a
used in verifying the A-46 equipment seismic adequacy at their plants. The staff expects,
l( these licensees to meet their commitments which the staff had accepted in
? the 120-day submittals. The staff's position is that licensees who deviated from their 120-( day in-structure response spectra commitments, should have specifically informed the l
L NRC of the change so the staff would have an opportunity to determine the adequacy of the revised spectra. The final determination of adequacy of individuallicensee's i responses will depend on the review outcome of each USl A-46 plant licensee's 1
2.
Operator Actions in Difficult Environmental Conditions NRC Evaluation in December 5.1996 Letter The NRC staff recognized that...the only events which must be considered (when performing the UEl A-46 program] are the SSE and loss of offsite power. However, each licensee should consider these factors on a case-by-case basis and determine what, if any, other complications they do need to address. During the August 28,1996, SQUG/NRC meeting, the staff provided SQUG with some additional clarification of the types of concems which should be considered including (1) the potential for diminished lighting due to loss of offsite power, (2) other baniers such as damaged equipment or structures which could inhibit operators ability to access plant equipment, and (3) the potential for requiring operators to enter hazardous or unfamiliar areas to manually reset or realign equipment.
Therefore, the staff does not consider a generic response to this question acceptable, and, on a plant-specific basis, each affected licensee should address this question as requested in the original individual plant RAl.
SQUG Response in April 18.1997 Letter During application of the GIP criteria and guidelines, SQUG members were advised that diminished lighting det to loss of offsite power should be considered when planning how.
to safely shutdown the plant following an earthquake. Some utilities have elected to add certain items of emergency lighting equipment to the safe shutdown equipment list.
3 Others have elected to rely upon emergency response tools, such as flashlights and harehold battery powered floodlights.
i l
NPC. L<1lw Nkl 3ln{92
& closure Q.q, y STAFF RESPONSE TO SEISMIC QUALIFICATION UTILITY GROUP CONCERNS i
- 1. Use of Ground Resoonse Soectra for Estimatino Seismic Demand The staffs RAfs to SQUG members have stated the staffs concern with the licensees' selection of demand spectra, have pointed out the cautions applicable in the event a licensee chooses to apply Method "A" of the GIP, and (in some cases) have requested the licensee to justify its selection of the demand spectra. In cases where a licensee elected to use different spectra than that provided by the licensee in its 120-day response to GL 87-02, Supplement No.1, the staff has requested the licensee to provide a basis for their decision. SQUG has taken the position that (1) a licensee submittal of the in-structure response spectrum (IRS) in its 120-day response did not imply a commitment to use the IRS exclusive of Method A, (2)
Method A is an altemative to Method B for elevations up to 40 feet from plant-grade level, (3) there is no requirement to use the more conservative of Method A or Method B. and (4) the GlP-2 cautionary paragraph (Pg 416 of GlP 2) is only to alert the seismic review team not to use Method A for atypical nuclear power plant structures.
f The staff does not agree with SQUG, for reasons previously provided in our letter of
' December 2,1997. The staff expects that the licensee must have a basis for the selection of a demand spectra and that the basis should include consideration of site specific factors )
'such as soil, the plant structure, the equipment location and natural frequency, and the 1
availability of an IRS. Site-specific situations cannot be disposed of solely on the basis of
)
generic arguments. The justification for the inclusion of Method A in the gip was that, for stiff structures such as those used in nuclear power plants, the amplification of the ground response spectrum up to the elevations of about 40 feet above plant grade levelwould be I
approximately 1.5 and that the original design bases IRS are overly conservative. We have found in our A-46 summary report reviews that the IRS developed specifically for this program using the latest modeling and analytical techniques can indicate amp _lifications R
greater thanA5 at,elegationslejo_w 4_0 feet.J owever, to move forward on the USI A-46 c!6sure prrcess, the staff intends to proceed as follows:
The staffis reviewing the use of Method A on a case-by-case basis and where a licensee provides an appropriate justification,its use will be acceptable.
Where a licensee's summary report indicates the use of a demand spectrum other than the licensing basis or that described in the 120-day response, the staff will continue to seek additional information regarding the methodology employed for developing the new spectrum, if the basis for the methodology was not provided.
Where a licensee's summary report indicates the selection of a different demand spectrum than that provided in the 120-day response, and sufficient information is provided regarding the methodology employed for developing the spectrum, the staff will not seek additional information on this issue.
If, in the staffs judgement, the licensee's use of Method A is not shown to be adequately justified, the staff will document this finding in the USI A-46 SER.
ENCLOSURE
MC La ffer bahl a(A c, Nclohre.
g EeF /
3 y
Staff's Evaluation of SOUG's Generic Response SQUG's response did not provide the requested.information. During the 28, 1996, SQUG/NRC meeting, the staff elaborated its concern and August the primary focus of the request for additional information (RAI) question. As e result of considerable discussions on the subject, the staff agreed to clarify the question. The following revised RAI will be forwarded to affected USl A-46 licensees for their response:
Referrina to the in-structure resconse soectra orovided in your 120-day-resconse to the NRC's reauest in Suoolement No. 1 to Generic Letter (GL) 87-02. dated May 22. 1992. the followina infomation is reauested:
Identify structure (s) which have in-structure response spectra (5%
a.
critical damping) for elevations within 40-feet above the effective grade, which are higher in amplitude than 1.5 times the SQUG Bounding
- Spectrum, With respect to the comparison of equipment seismic capacity and b.
seismic demand, indicate which method in Table 4-1 of GIP-2 was used to evaluate the seismic adequacy for equipment installed on the corresponding floors in the structure (s) identified in Item (a) above.
If you have elected to use method A in Table 4-1 of the GIP-2, provide a technical justification for not using the in-structure response spectra provided in your 120-day-response.
It appears that some A-46 licensees are making an incorrect comparison between their plant's safe shutdown earthquake (SSE) ground motion response spectrum and the SQUG Bounding Spectrum. The SSE ground motion response spectrum for most nuclear power plants is defined at the plant foundation level. The SQUG Bounding Spectrum is defined at the free field ground surface. For plants located at deep soil or rock sites, there may not be a significant difference between the ground motion amplitudes _ at the foundation level and those at the around surface. Slowever, for siTifs where a structure is ~ founded _on sha ow 5 amplification of the ground motion from the foundation solevel to the ground surface may be significant y For the s' ucture(s) identified in Item (a) above, provide the c.
in-strum -
response spectra designated according to the height above tu ifective grade. If the in-structure response spectra identifi in the 120-day-response to Supplement No. I to GL 87-02 i
was not used, provide the response spectra that were actually used to 1
verify the seismic adequacy of equipment within the structures
{
identified in Item (a) above. Also, provide a comparison of these
{
t 1
spectra to 1.5 times the Bounding Spectrum.
The staff does not consider a generic response to this question acceptable, and affected licensees should address this question on a plant-specific basis.
l l
L__
' 6 Q MG Le & r Nd c fso(qy G lorure. i W 3/
SQUG Response to NRC Statement in August 6,1996 letter O.-
'Ihe basis for requiring affected licensees to review the seismic adequacy of mechanical and electrical equipment is contained in Generic letter (GL) 87-02 (February 19,1987) and its supporting references NUREG-1030 and NUREG-1211. In GL 87-02, the NRC concluded that
...the seismic adequacy of certain equipment in operating nuclear power plants must be reviewed against seismic criteria not in use when ibeie olants were licensed.
[ Emphasis added]
l The NRC also recognized that
^
Direct application of current seismic criteria to older plants would require extensive, and probably impractical, modification of these facilities.
Therefore the NRC endorsed an alternate resolution of this problem in the enclosure to the Generic Letter which is based upon use of earthquake experience data supplemented by test results. The enclosure to GL 87-02 explicitly allows licensees to use the " bounding spectra" defined by the Senior Seismic Review and Advisory Panel (SSRAP) as the basis 3
for evaluating the seismic adequacy of equipment rather than require equipment capacity to be compared to floor response spectra:
.The purpose of these bounding spectra is to compare potential seismic exposure of equipment in a nuclear power plant with the estimated ground motion that similar equipment actually resisted in earthquakes described in the [ earthquake experience] data base.
For convenience, the bounding spectra are expressed in terms of ground response at the nuclear site rather than floor response or equipment response.
[GL 87-02. Enclosure. Page 8]
As a result of additional work completed by SQUG and SSRAP after OL 87-02 was published, a refm' ed " Bounding Spectrum" and procedure for comparing the Bounding Spectrum to a plant's safe shutdown earthquake (SSE) was defined and included in the Generic Implementation Procedure (GIP) as Method A. The technicaljustification for using this Bounding Spectrum and the associated procedure for evaluating equipment
. seismic adequacy is documented in the SSRAP report (Reference 3).
11 ____ - __ __ _ __-__ _ -_ _
~
souc c 4 w, u,
E SQUG
... =
l November 11,1997 Mr. John F. Stoh U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulations 11555 Rockville Pike - Mail Stop 14D07 Rockville, MD 20555
Subject:
Seismic Qualification Utility Group (SQUG)
Clarification of GIP Provisions on Use of Method A
Dear Mr. Stolz:
During recent meetings with the NRC staff the subject of seismic demand and the use of GIP Method A have been discussed. Some of the statements made by representatives of the NRC staff can be interpreted to indicate they may consider that spectra reviewed by the NRC in tlie utilities' 120-day response letters to Supplement No.1 to Generic Letter (GL) 87-02, are reouired to be used for all A-46 evaluations, effectively eliminating the option of Method A. To clarify SQUG's position, the following information is provided for NRC consideration.
1.
GIP-2, approved by SSER #2, provides the following options for seismic demand:
Section 4.2 of the GIP states:
"ne seismic capacity of an item of equipment can be compared to a seismic demand which is defined in terms of either a ground response spectrum nr an in-structure response spectrum. Table 4-1 outlines three types of comparisons as either Method A or B." (underline added)
Section 4.2.3 of the GIP further elaborates:
"The ground response spectrum for the Safe Shutdown Earthquake (SSE) can be used to represent the seismic demand applied to nuclear plant equipment when one of the following two comparisons is made:
Method A.1:
Bounding Spectrum envelops the SSE ground response spectrum (5% damping).
Method A.2:
GERS envelops 1.5 times 1.5 times the SSE ground response spectrum (5 damping)."
.s
L 69 LAG LeCddd afh(9 r Mr. John F. Stolz November 11,1997 Finally, Section 4.2.3 of the GIP makes clear that the use of ground response spectra (Method A) in lieu of in-structure response spectra (Method B) is not only for those plants which have no design in-structure response spectra:
l l
"'The advantage of using ground response comparisons is that with the applicable restrictions and limitations, all the equipment covered by the Bounding Spectrum or the l
GERS can be evaluated for seismic adequacy without the need for using in-structure response spectra which are eften based on very conservative modeling techniques or may not be available."
2.
. Because judgment is required to classify spectra as " conservative" or " median-centered,"
NRC requested certain licensees to submit descriptions of spectra so that the NRC could determine if they are " conservative" or " median-centered." (Some utilities which used more modern methods were advised they could consider their spectra as
" conservative" without any spectra reviews by the NRC.) Licensees complied and the NRC reviewed and designated the spectra as " conservative" or " median-centered" on a case-by-case basis.
3l Neither the NRC requests nor responses in any way stated that these spectra were the only seismic demand options which could be used, nor did they change the seismic demand options approved by SSER #2 in the GIP.
Therefore, SQUG's position is that the Method A option continues to be available (provided l
special conditions are met, i.e., the equipment is mounted below about 40 feet above the effective grade and its lowest natural frequency is above 8 Hz) based on the clear rules in the GIP and it is applied properly as clari5ed in SQUG's letter of June 30,1997 to the NRC on this subject. Any attempt to require use ofDaly the 120-day spectra would represent a change in the NRC's published position and would have to be justified under 10 CFR 50.109.
Sincerely,
/
Neil P. Smith, Chairman Seismic Qualification Utility Group
.cc G. C. Lainas, NRC, MS O-7D26 D. H. Dorman, NRC, MS O-14C1 R. Wessman, NRC, MS O-7E23 K. Manoly, NRC, MS O-7E23 P. Y. Chen, NRC, MS O-7E23 R. P. Kassawara, EPRI SQUG Steering Group SQUG Representatives i
J
SQuG L<$r Nr/IltWAY Mr. John F. Stolz January 26,1998 SQUG's understanding of the NRC positions in Reference 1 concerning use of Method A is that the utilities are bound to the following two conclusions: (1) when SQUG utilities provided information in their 120-day response letter (as requested by the staff in Supplement No.1 to GL 87-02 and SSER No. 2) on the methods used to calculate ISRS, licensees committed to using those spectra, and only those spectra, for -
all GIP evaluations and (2) that a precautionary statement in the GIP precludes use of Method A any time the plant's ISRS exceed the Method A spectra. These NRC conclusions are incorrect, inconsistent with the approved GIP, and differ from the understanding among SQUG, the independent expert review panel, SSRAP, and the NRC staff at the time the GIP was developed. Our description of these two issues is provided below:
I a.
Pumose for Submitting ISRS to NRC l
i As explained in SQUG's letter of November 11,1997, some SQUG members y
submitted information on methods used to calculate ISRS so the NRC could I
I agree on whether they were " conservative, design" or " realistic median "
centered." Neither the information request nor the responses in any way involved the commitment now being inferred by the staff. Further, no changes were made to the GIP provisions which give utilities the option to use gilhct Method A at Method B.
It is noteworthy that the NRC did not request ISRS information from some l
SQUG members (e.g., SEP plants) because they had previously reviewed them L
on other projects or already had the ISRS information in their files. With this previously obtained information, the NRC concluded in SSER No. 2 (page 14) that the ISRS in the licensing basis documents of these 15 plants "may be used as ' conservative, design'in-structure response spectra." SSER No. 2 does not require these plants to use ISRS (Method B) exclusively nor does it draw a distinction between these plants and the other USI A-46 plants, except with respect to whether the ISRS should be classified as " conservative, design" or
" realistic, median-centered."
There is nothing in GIP-2 or SSER No. 2 which limits use of the GIP to Method B. GIP-2 clearly states,in Part II, Section 4.2, that either Method A or Method B may be used to compare seismic capacity to seismic demand:
"The seismic capacity of an item of equipment can be compared to a seismic demand which is defined in terms of either a ground response spectrum or an in structure response spectrum. Table 4-1 outlines these types of comparisons as either Method A or B. Method A is for making a comparison with a ground response spectrum; Section 4.2.3 discusses this type of comparison. Method B is a comparison with an in-structure a
S&vG h 6 %6plufu(qy Eefs.6/
Mr. John F. Stolz January 26,1998 h
response spectrum; Section 4.2.4 discusses this type of comparison.
Method A comparisons are generally easier to apply than Method B t
comparisons."
(GIP-2, pg. 4-8)
SSER No. 2 does not take exception to this provision in the GIP but concludes that the screening procedures and criteria are adequate and acceptable:
/
"The staff has reviewed the screening procedures and criteria. Based on
/
the evaluations and findings described in Sections 11.4.2, II.4.3, and II.4.4 I
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, I
exceptions and positions described in the sections that follow."
(SSER No. 2, pg.12)
There are no staff clarifications, interpretations, exceptions, or positions in SSER No. 2 which contradict our conclusion that licensees may use either Method A or Method B.
J l
The above interpretation of NRC's position on GIP-2, as contained in SSER No. 2, is supported by the NRC position in GL 87-02 which clearly allows use of 1.5 times ground response spectra (later called GIP Method A)in lieu of plant floor response spectra (also called ISRS):
"When horizontal floor spectra exist, these spectra may be used to obtain the equipment spectral acceleration. Alternatively, for equipment f
mounted less than about 40 feet above grade, one-and-a-half times the i
free-field horizontal design ground spectrum may be used to conservatively estimate the equipment spectral acceleration. For equipment mounted more than about 40 feet above grade, floor spectra (GL 87-02, Enclosure, Section 3, page 5)
Therefore, we conclude that the NRC's request for information in GL 87-02, Supplement No. I and SSER No. 2, and their review of these LB ISRS, was for l
the purpose of evaluating the adequacy of these spectra for use in GIP Method l
l B and for deciding whether to classify them as either " conservative, design" or
" realistic, median-centered," The licensee submittal of the ISRS in the 120-day L
response does not imply a commitment to use these ISRS exclusive of i
l Method A. GIP Method A is clearly an alternative to Method B, and no I
requirement to use the more conservative of Method A or Method B is stated or I
implied.
h l
t-m
._=
SQu(> La fLr- %fe/ s hoh, M S Mr. John F. Stolz
.2-June 30,1997
,.s reiterates that GIP Method A has been previously reviewed and accepted by recognized experts, including the NRC staff, as an acceptable ethod to verify the seismic adequacy of equipment installed in operating nuclear plants. The on y issue SQUG is raware of thatIhe NRC has identified in the application oTMethod A is the proper consideration of the location of the safe shutdown earthquake (SSE) ground response spectra. In Enclosure 2, SQUG clarifies the issue regarding the location of the SSE ground response spectra and where it is typically defined for most USI A-46 plants. Also a set of screens has been suggested by SQUG for its members to use when responding to the J
subject RAI question. These screens help members respond to the area where the NRC believes GIP Method A may be misapplied, i.e., for sites where a structure is founded in
} shallow soil with the location of the SSE ground response spectra defined below the top of i
the ground surface. Other than this clarification, SQUG is not aware of any other NRC concerns relating to the proper application of GIP Method A for resoMng USI A-46.
If the NRC has any new information that could bring into question the use of GIP Method A, please provide the information to us so SQUG can evaluate whether the GIP should be revised. Otherwise SQUG considers GIP Method A an approved method for evaluating seismic capacity compared to demand when the restrictions provided in are met.
5 i
Sincerely,
)
i Neil P. Smith, Chairman Seismic Qualification Utility Group Enclosures (2) cc:
D. H. Dorman, NRC, MS: 0-14C7 R. Wessman, NRC, MS: O-7E23
- K. Manoly, NRC, MS: 0-7E23 P. Y. Chen, NRC, MS: O-7E23 R. P. K.tssawara, EPRI -
SQUG Steering Group SQUG Representatives and Alternates i
i I
I I
i Copies of References Associated With issue #2 Justification for Use of Cable Tray Methodology (Documents Arranged in Order of Reference Number) l i
I l
ENCIASURE 4
Ref.1/
6IN Revision 2 Corrected,6/28/91 Floor-Mounted Sunoorts.
Plastic behavior of floor-mounted supports' may lead to structural instability. Ductility, as defined by these guidelines, only. applies to suspended systems.
Floor-mounted supports are characterized as non-ductile, and are subject to further horizontal l strength review in Sections 8.3/6 and 8.3.7 with focus on stability.
Rod Hanaer Traneze Suonorts.
Supports constructed of threaded steel rods with fixed-end connection details at the ends of the rods behave in a ductile manner under horizontal motion; however, relatively short rods may undergo very large strains due to bending imposed by horizontal seismic
. motion, at'the' fixed ends of-the rods. Low cycle fatigue may govern response.
Rod hanger trapeze supports with short, fixed-end rods should be j evaluated for low cycle fatigue effects in Section 8.3.5.
No further review of horizontal response capability is required of supports characterized as ductile. Only the support vertical capacity need be j verified, as discussed in Sections 8.3.1 and 8.3.2.
If a support is characterized as non-ductile or has questionable ductility, then its lateral load capacity should be verified, as discussed in Section 8.3.4, as shown in the logic diagram for making these decisions in Figure 8-6.
8.3.4 Lateral Load Check A Lateral Load Check should be performed for the bounding case raceway supports that are characterized as potentially non-ductile. The Lateral Load Check is in the form of an equivalent static lateral load coefficient.
The Lateral Load Check compares the ratio of horizontal load capacity divided by dead load demand (for potentially non-ductile supports) to the same ratios for support systems in the seismic experience data base that performed well.
Because many of these data bas'e raceway systems were subjected to earthquake ground motions that may have been greater than the Safe Shutdown Earthquakes (SSEs) for many plants, provisions for scaling down the equivalent static horizontal. loads are given below.
a support is ductile, then no further review of horizontal response I capability is required, and the support may be shown to be seismically
-l gged by the Vertical Capacity. Check Section 8.3.2)[If a support is 8-27
Rd )
3 g p._ L Revision 2 Corrected,6/28/91 The 3.0 times dead load static coefficient should not be reduced if the design basis earthquake ground motion response spectrum for that plant site is less than the Bounding Spectrum shown in Figure 4-2.
This is because there are only a few supports in the earthquake experience data base which l have back-calculated vertical capacities less than 3.0 times Dead Load.
If l the 3.0 times Dead Load guideline is not met, then the support should be l classified as an outlier. Resolution of the outlier can be accomplished by l the methods described in Section 8.4.
l 8.3.3 Ductility Check An evaluation should be conducted of the supports selected for review to characterize their response to lateral seismic motion _as either ductil_e or potentially non-ductilepSupports suspended only from overhead may be characterized as ductile if they can respond to lateral seismic motion by swinging freely without degradation of primary vertical support connections and anchorage. Ductile, inelastic performance such as clip angle yielding
(~S or vertical support member yielding is acceptable so long as deformation t lead to brittle or premature failure of overhead vertical support.
Review of typical conduit and cable tray support systems in the earthquake experience and shake table test data bases indicates that many overhead mounted support types are inherently ductile for lateral seismic motion.
Back-analysis of many data base conduit and cable tray supports predicts yielding of members and connections. These data base systems performed well, with no visible signs of distress.
Ductile yielding of suspended supports results in a stable, damped swaying response mode.
This is considered to be acceptable seismic response.
l 8-24
SS EE No,'2, (2t[4 L 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 deadicad (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 C
i cable trays and conduit raceway in the US! A-46 plants is as strong as those D
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 US! 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 acceptades 3
g approach for evaluating the seismic adequacy of existing cable and conduit g raceways in USI A-46 plants. Also, the staff agrees that the proposed L
../
I i
SSec No. 2., r/n/qt Raf. 1 h 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 Although the proposed guidelines would not require detailed analyses ' data. 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. l Therefore, the staff concludes that the proposed guidilines for evaluation of 1 seismic adequacy of cable and. conduit raceways and th9ir supports are Qceptable subject to the staff evaluations described in this supplement._ ~ ~- II.9 Documentation Section 9 of Part !! 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 a 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. 11.10 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 US! A-46 program, they must be submitted for staff review and approval. 31 g.
f d C, L e N r-Ned 12,[B[41 belosurc 24f, p The analysis checks described in the GIP process are intended to confirm that nuclear plant cable tray systems are bounded by the earthquake and test experience. They are not intended to satisfy typical structural analysis criteria.
NRC Staff Evaluation
l The staffs concem is with the concept that the cable tray and conduit supports in the experience data base performed well because they wer e ductile and the fact that without any quantitative analysis of the ductility of the supports, the criteria is extended to A-46 plants. On several occasions SQUG has indicated that it would provide information on ' the basis for the criteria but, to date, it has not done so. In view of this, the staff will ask licensees of A-46 plants to provide the following plant-specific information about the evaluation of cable tray and raceway supports using the ductility concept. Seismic Adequacy of Cable and Conduit Raceways The NRC staff has concems about the way the A-46 cable trays and conduit raceways issue was being dispositioned by licensees. We issued RAI to several licensees on this issue. SQUG responded instead of the licensee because SQUG considered the RAI to be generic in nature. The staffissued a subsequent RAI to SQUG as a follow up to their response. However, the staff found that the correspondence with SQUG did not achieve the intended results in that they did not address the technical concems of the staff. Therefore, we are restating our concems in the following discussion. The GIP procedure recommended performing what is called a limited analytic evaluation for selected raceways and cable trays. The procedure further recommended that when a certain cable tray system can be judged to be ductile and if the vertical load capacity of the anchorage can be established by a load check using three times the dead weight, no further evaluation is needed to demonstrate lateral resistance to vibration from earthquakes. The staff has concems with the manner in which these simplified GIP criteria were implemented at your plant. The GIP procedure eliminates horizontal force evaluations by invoking ductility. However, all the so called non-ductile cable trays would eventually become ductile by inelastic deformation, buckling or failure of the non-ductile cable tray supports and members. If this procedure was followed for eliminating cable trays for further assessment at your plant, then all the cable trays could conceivably be screened out from A-46 evaluation. We are requesting your response on the following items to elicit information that would support our safety evaluation of cable trays at your plant. 1. Provide the total number of raceways that were classified as ductile in your A-4S evaluation and for which you did not perform a horizontal load evaluation. Indicate the approximate percentage of such raceways as compared with the entire population of raceways. Discuss how the ductility concept is used in your walkdown procedures. Provide descriptions of all the typical configurations of your betile raceways) 2. (dimension, member size, supports, etc.)f g ~ Provide justification for stating that_ ductile _ raceways need_not be evaluated fo) 3. horizontaUoad,rwhen a reference is provided, state the page number and paraghiiphs. The reference should be self contained, and not refer to another
NSC Lder ba+d g(qq [ br.l seee O g \\ reference. You should siso provide a technical discussion why the reference you provided constitutes bases for the ductility methodology. 4. In the evaluation of the cable trays, and raceways, if the ductility of the attachments is assumed in one horizontal direction, does it necessarily follow that the same system is ductile in the perpendicular direction? frovid_e a deGnition of ductility in engineeri_ng terms [m vide an assessment of 5. ( the maximum ductility utilize for the weakest cable tray support. . 6. Discuss raceways and cable trays that are outside of the experience data by explaining criteria used for making your safety determination, the configurations of such raceways and the number of the raceways and percentage with respect to the whole population of raceways. How are they going to be evaluated and disposed? The staff needs this information to make its finding as to the adequacy of the information provided in response to Gt. 87-02. 6. }_eismic Adeauacy of Relays Mounted on Diesel Generators and Air Compressors 1 NRC Evaluation in December 5.1996 Letter The SQUG's generic response to the NRC question is not acceptable. First, the specific issue in questien is not regarding relays mounted on diesel generators and air compressors. The issue is the inappropriateness of using the "mie-of-the-box" concept and the judgment based on the normal operation of the diesel gcaerators or air compressors to justify the seismic adequacy of devices, such as relays, mounted in the instrumentation and control cabinets anchored on the common skid of the Csel generator or air compressor. The concept of the " rule-of-the-box" applies to components in a system that has already been successfully subjected to a vibratory environment comparable with or greater than the required motion (e.g., SSE). Therefore, the " rule-of-the-box" concept can also be applicable for acceptance of the relays mounted on vibratory equipment (or in a cabinet supported on the common skid) provided it is demonstrated that the vibratory motion of the equipment (or the skid) is at least equal to the required seismic motion at that location, and that the relays performed all their intended functions during the periods of vibration. With regard to relays mounted on diesel generators and air compressors the following specific questions should be addressed: a. Does the mechanical vibration envelop the required input motion (e.g., SSE) from all aspects (e.g., amplitude, frequency, direction, etc.)? b. Do these vibratory equipment-mounted relays perform all their operational safety functions (e.g., change of state) during the mechanical vibrations (i.e., during startup and normal operation) so that the relays can be considered qualified to that level?
' ~ Md (.4 hs-Abd, 3[E 7N G ^c ' ** " " %5, y 1
- 2. t_ateral Lead Ductility Evaluation of Cable Travs On a site-specific basis, the staff has questioned the licensee's application of the GIP l
methodology with regard to the assumption of ductility of cable tray and raceway supports. l The staff has questioned the basis for performing a limited analytical review and has asked various plant-specific questions regarding the evaluation of cable trays. SQUG believes certain of the staff questions are generic in nature and provided with its January 22,1998, letter, additional materialin support of the GIP-2 guidelines for the venfication of cable tray seismic adequacy. The staff reviewed and accepted the approach provided in GIP-2. The staff concem is not necessarily with the GIP-2 concept. but rather with the way it has been employed by some licenseesjonsequentl9, licensees have been asked to explain their bases for etefminations made during plant walk downs that cable tray and raceway supports are f(ductile. The staff accepts SQUG's position that cable tray and receway supports in th
- base are inherently rugged because they generally behaved in a ductile manner. However, l for A-46 plants with cable tray or raceway configurations which are not in the experience i database or those with obviously rigid support configurations the staff will continue to ask plant specific questions regarding a licensee's basis _for their ductility determinations hese p ant specific aspects can'not'be addressed genencally.
~
- 3. Seismic Adecuacy of Refsvs Mounted on Diesel Generators and Air Compressors At one facility (Point Beach) the staff requested the licensee to justify its basis for relay operability for cer1ain relays mounted in control cabinets anchored on a common skid with a diesel generator or air compressor. SQUG assumed this was a generic issue and expressed concern that the staff was taking issue with the provision to accept relays mounted on equipment such as diesel generators and air compressors.
This original question was focused at relays in control cabinets anchored to a common skid with the diesel generator or air compressor, as opposed to relays mounted directly on the equipment. This issue has not been identified at any other facility and the Point Beach licensee has agreed in its response of July 31,1997, to reexamine the staff's original concem. The staff believes this. issue is closed.
- 4. Justification for Schedules to Resolve GIP Outliers Staff RAls have asked licensees about their schedules to resolve indeterminate outliers and have inquired if licensees have addressed possible operability issues. SQUG stated in its letter of January 26,1998, that licensees are committed to perform a seismic adequacy review and report the results to the NRC in response to the NRC's 50.54(f)information request. SQUG reminds NRC that, as stated in NUREG-1211 (Regulatory Analysis for Resolution of Unresolved Safety issue A-46, Seismic Qualification of Equipment in Operating Plants, February 1987) a backfit analysis for the correction of any deficiency will be performed on a case-by-case basis if required.
SQtg Lefkr 'pdeci tht(4YRJ. y Mr. John F. Stolz January 22,1998 itself, or to provide additional evaluations beyond those required by the GIP, even though the NRC accepted the GIP as being adequate for evaluating raceway systems for resolution of USI A-46 in SSER No. 2.
- --%.=.-=,
Tor example, Part 3 of the RAI question in Enclosure 1 asks licensees to " provide N f' justification for stating that ductile raceways need not be eval The GIP requires a lateralload evaluation of only those raceway supports that can be I characterized as potentially non-ductile. The GIP explicitly exempts ductile raceway , supports from being reviewed for horizontal response capability. f "If a support is ductile, then no further review of horizontal response capability is required, and the support n:ay be shown to be seismically rugged by the Vertical ; Capacity Check (Section 8.3.2)." j i (GIP-2, Section 11.8.3.4) ) l The basis for this approach is discussed in the enclosures to this lette fan example $here the NRC asks SQUG members to provide information beyond the scope of the GIP is in Part 5 of the RAI question in Enclosure 1; the licensee is asked to provide an assessment of the maximum ductility utilized for the weakest cable tray y q{ " support." The methodology contained in Section 8 of the GIP ) i I evaluation of the " maximum ductility" for any of the cable tray supports. The effect of i i j support ductility has been implicitly accounted for in the GIP methodology. This I request for additional evaluations appears to be beyond that required by the GIP. Therefore, this request appears to represent a change in the NRC's previous position ~ __ f-l which would have to be justified under 10 CFR 50.109. During discussions between Neil Smith (SQUG Chairman) and Richard Wessman (NRC/NRR),it was agreed that it would be helpful to current members of the NRC l staff, who were not involved in reviewing the GIP and preparing SSER No. 2, to see the content and extent of the earthquake experience database. Therefore, we arranged for j members of the staff to visit EQE Internationalin California in October 1997. They l also had the opportunity to visit one of the database sites (Moss Landing Power Plant) which had been subject to the 1.cma Prieta earthquake (Magnitude 7.1)in 1987. During this trip, they inspected equipment, including cable and conduit raceway systems, which had successfully performed their intended function during an earthquake larger than the Safe Shutdown Earthquakes (SSEs) for eastern U.S. nu& ar plants. This visit was similar to numerous other meetings and onsite inspections performed by members of the NRC staff during the 1980s when the GIP and its reference documents wer: reviewed and approved. 1 I 1 L_______
] SQ\\AG Leike Ne/ IlatNV (M..y Mr. John F. Stolz January 22,1998 Some of the questions and subjects covered in the RAI of Reference 1 are the same as the issues discussed and resolved when the cable and conduit raceway criteria and guidelines were being reviewed by the NRC staff during the period from 1987 to 1992, l i.e., prior to issuance of SSER No. 2. As indicated in our June 11,1997 letter to you (Reference 2), we do not consider it necessary or warranted to re review the basis for the GIP criteria and guidelines which were accepted by the NRC staff in SSER No. 2. To our knowledge, new information has not been identified which invalidates the previous staff position that the GIP is an acceptable method for verifying the seismic adequacy of cable and conduit raceways. ) We understand that the ERC staff who are currently reviewing utility implementation of the cable tray evaluations were not party to the discussions and meetings held between NRC, SSRAP, and SQUG prior to the NRC's acceptance of the GIP in SSER No. 2. We also 'mderstand that some of the documentation describing the basis for the cable tray evaluation criteria and procedure may not be available_irl.thh f Therefore, to help the current members of the staff better understand the basis for G criteria and guidelines which were reviewed and approved in SSER No. 2, the following three enclosures are included for your use. The first item (Enclosure 2) is an SSRAP ( report sent to the NRC in 1987. The.other two enclosures summarize some of the material available during the original review of the GIP and show how it supports the / GIP criteria and guidelines. f is a copy of a report which was originally sent on November 13,1987, l I to the NRC staff by the Senior Seismic Review and Advisory Panel (SSRAP). This material was thoroughlydiscussed and evaluated during joint meetings between the NRC staff,SSRAF, and SQUG and forms one of the bases for not / performing lateralload checks on ductile raceway supports. This enclosure ] includes a summary of the results of calculations performed on 54 support j { calculations from the earthquake experience database. It also calculates the j l margin factors for several of these supports to calibrate the factors chosen for the i limi:ed analytical review contained in the GIP. ) ' is a report which summarizes the basis for the cable tray methodology contained in the GIP. It also summarizes the results of numerous tests programs which were available to the NRC staff during their review of the GIP from 1987 to ( 1992. ( is a report prepared for SQUG by the chairman of SSRAP, Dr. Robert P. Kennedy, which supplements the material contained in Enclosures 2 and 3. It offers additional details on the earthquake and test data ) and summarizes SSRAP's conclusion that it is not necessary to perform an explicit lateralload evaluation of cable tray supports which are judged to be ductile. Q
SQW;Ledy h fed //u/gt I/ b d o t u v e. 1 "An evaluation should be conducted of the supports selected for review to l characterize their response to lateral seismic motion as either ductile or l l potentially non-ductile. Supports suspended only from overhead may be characterized as ductile if they can respond to lateral seismic motion by swinging freely without degradation of primary vertical support connections and anchorage. Ductile, inelastic performance such as clip angle yielding or vertical support member yielding is acceptable so long as deformation does not lead to brittle or I premature failure of overhead vertical support.' (GIP-2, Part II, pg. 8-24) Each of the worst case samples of raceway supports which were selected for the GIP Limited Analytical Review (LAR) were evaluated using the procedure given in GIP Section 11.8.3.3 to determine whether they are ductile or non-ductile based on the examples shown in GIP Figures 8-7 and 8-8. The GIP methodology does not require an evaluation of the " maximum ductility" for any of the cable tray supports. We consider our review of the effect of ductility to be in accordance with GIP-2, as approved by SSER No. 2, which implicitly accounts for ductility in its derivation. Since this RAI is a request for additional evaluations beyond that required by the GIP,it appears to represent a change in the NRC's previous position which would have to be justified under 10 CFR 50.109. NRC RAI Ouestion Part 6 6. Discuss raceways and cable trays that are outside of the experience data by explaining criteria usedfor makingyoursafety determination, the configurations ofsuch raceways and the number of the raceways andpercentage with respect to the whole population of raceways. How are they going to be evaluated and disposed? Suppested Resoonte All the cable and conduit raceway systems in the plant are within the scope of the seismic review procedures contained in the GIP except those which were identified as ( outliers. GIP Section 11.8.0, page 8-2, paragraph 2 describes the scope of raceway I systems which are covered by the GIP as follows: I i } "The seismic review guidelines contained in this section are applicable to steel and aluminum cable tray and conduit support systems at any elevation in a nuclear power plant, provided the Bounding Spectrum (shown in Section 4, Figure 4-2) envelopes the largest hori7ontal component of the 5% damped, free-field, safe shutdown earthquake (SSE) ground response spectrum to which the nuclear plant is licensed." (GIP-2, pg. 8-2). _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
SQ4 Leffer lb4d e/z2fy 6tio3vre$ M.E.J The raceways which were classified as outliers did not meet the " Inclusion Rules," had / "Significant Other Seismic Performance Concerns," or did not satisfy the " Limited Analytical Review Guidelines" contained in the GIP. The resolution of outliers may include additional analysis, investigation, or a modification to the raceway which would I allow the raceway to meet the GIP screening criteria. q [If not provided in the p ant's A-46 Summary Report, describe the cable and conduit system outliers identified during the USI A-46 review and describe how these outliers have been or will be resolved.) The RAI question also asks for the percentage of raceways which are outliers with respect to the whole population of raceways. As described in the response to Part 1 of this RAI question, we did not collect this information during the raceway walkdown since it is not part of the GIP guidelines. For the reasons cited in response to Part 1 of this RAI question, we request that NRC staff reconsider the need for establishing the percentage of outlier raceways compared to the whole population of raceways in the plant. REFERENCES 1. NRC (J. Stolz) letter to SQUG (N. Smith) dated December 2,1997, Evaluation of Seirmir Qualification Utility Group's Response to Generic Issues Included in NRC's Request for AdditionalInformation. 1 2. SSRAP Report," Review Procedure to Assess Seismic Ruggedness of Cantilever { Bracket Cable Tray Supports," Senior Seismic Review and Advisorf anel (SSRAP), j P Rev. 3.0, March 1,1991. 1 ) l I I l 1 ) \\
4 Copies of References Associated With issue #3 Adoption of GlP as a Licensing Basis Method (Documents Arranged in Order of Reference Number) l l 1 1 I i DCIDSURE 5 l 1 L___ 1
Q.23-O'L,kt4f87 belosure M,!/ For components included in the data base by type but outside the Ifmits of ex-or of a type not included in the data base as a perience data or test data generalguidelinetheseismIcverificationcanbedeferreduntiladditionaltest data is developed, endorsed by SSRAP, and approved by the NRC staff, provided that the seismic verification is completed no later than about 36 months from the date of issuance of the USI A-46 final resolution. Actual schedule dates l will be based on negotiations with the generic group or with individual utili-ties. The proper integration of the proposed work scope into each plant's schedule for plant modifications will be considered. fiTautilityrep1ac'escomponentsfiranfreason each replacement (assembly, ~ ]criteriaandmethodsorsubassembly, device)mustbeverifiedforseism as an option, qualifying by current licensing criteria. Thisprovisionalsoappiiestofuturemodificationorreplacements. " Component", 4-- - in this context means equipment and assemblies (including anchorages and sup-ports)--suchaspumpsandmotorcontrolcenters--andsubassembifesanddevices--) such as motors and relays that are part of assemblies. _ 3. Verification of Anchorage To verify acceptable seismic performance, adequate engineered anchorage must be provided. There are numerous examples of equipment sliding or overturning under seismic loading because anchorage was absent or inadequate. Inadequate anchorage can include short, loose, weak, or poorly installed bolts or expansion anchors; inadequate torque on bolts; and improper welding or bending of sheet metal frames at anchors. Torque on bolts can normally be ensured by a preventive maintenance and inspection program. In general, checking of equipment anchorages requires the estimation of equipment Also, one must either estimate weight and its approximate center of gravity. the fundamental frequency of the equipment to obtain the spectral acceleration at this frequency or else use the highest spectral acceleration for all fre-quencies. When horizontal floor spectra exist, these spectra may be used to obtain the equipment spectral acceleration. Alternatively, for equipment mounted less than about 40 feet above grade, one-and-a-half times the free-field hori-zontal design ground spectrum may be used to conservatively estimate the equip-ment spectral acceleration. For equipment mounted more than about 40 feet above grade, floor spectra must be used. This restriction may be modified if addi-tional data become available to justify raising the 40-fcot-limit. Equipment anchorage must not only be strong enough to resist the anticipated forces but must also be stiff enough to prevent excessive movement of the equip-ment and potential resonant response with the supporting structure. The review of anchorages should include consideration of both strength and stiffness of the anchorage and its cciponent parts. Additional discussions on seismic motion bounds and equipment supports and anchori;ge for each of the original eight classes of equipment in the experience data base are included in paragraph 6 below. This guidance supplements the general guidance above. During the walk-through inspection anchors and supports of equipment within the scope of review will be carefully inspected. The detailed guidance devel-oped is the preferred method for review of anchorages. The detailed guidance has been developed jointly by SQUG and EPRI. It was approved by SSRAP and is 5 Enciosure
GIP-i, Peti 2*f II Revision 2 Corrected,2/14/92 satisfy the purpose of the NRC regulations relevant to equipment seismic adequacy (including 10 C.F.R. Part 100) and applicable to the licensee's l plants for the identified equipment.' 2.3.3 Revision of Plant Licensina Basti USI A-46 licensee, in accordance with 10 C.F.R. s 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.fTicensees should include deviations from the SQUG commitments and guidance of the GIP in their unreviewed safety question analyses under 10 C.F.R. s 50.59. (See Example 5 and Paragraph 2.3.4, below, for extending the scope of plant equipment to which the revised licensing bases apply.) With the exception of cable and conduit raceway systems, the USI A-46 (GIP) methodology is an equipment-level seismic verification methodology for mechanical and electrical equipment, tanks and heat exchangers. For A-46 plants, this methodology is adequate to verify the seismic adequacy of equipment within the scope of the GIP. However, it is recognized that in some cases the plant licensing basis may address aspects of the seismic adequacy of the systems, in which this equipment is installed, that are not addressed by the GIP. Where this is true, licensees are not relieved of the responsibility to esaluate and adequately address scch system-level seismic requirements as part of the revision of.a plants' licensing bases for equipment. The USI A-46 (GIP) methodology shall not supersede any seismic qualification requirements imposed or comitted to in connection with the resolution of other specific issues (e.g., Regulatory Guide 1.97, TMI Action Item II.F.2, and Individual Plant Examination for External Events) 12 ../
Gip-z.,Pa,t1 MN ^ Revision 2 Corrected,2/14/92 unless these qualification requirements or commitments are also revised l according to appropriate regulatory requirements, where applicable.5 To help clarify the intent of this section for modifyir.g the licensing bases of the plant, the following examples are provided. These examples explore some, but not all, of the possible scenarios that may be encountered by licensees when revising their licensing bases to adopt the USI A-46 (GIP) methodology (or an alternative) as the method for verifying the seismic adequacy of electrical and mechanical equipment within the i scope of equipment covered by the GIP. 3 Examole 1 ) Revising the plant licensing bases when the plant is not currently ( committed to using any specific method to verify the' seismic adequacy ) of equipment, and there are no specific commitments to seismic /N qualification requirements for equipment connected with the resolution () of other specific issues. I When a plant has no general commitment to methods or standards for seismic verification or qualification of equipment, i.e., the i FSAR is silent, and no specific licensing commitments exist for specific issues (as discussed in Example 2, below, the plant may adopt the USI A-46 (GIP) methodology without specific notification of the NRC. This is neither a license change nor a change to the facility as described in the FSAR. Nevertheless as with any change in the plant procedures or methodology for evaluation of plant adequacy, the basis for the change should be l documented. One option available is a safety evaluation pursuant to 10 C.F.R. 5 50.59 together with a formal FSAR change (if appropriate) in accordance with 10 C.F.R. s 50.71(e). Examole 2 i Revising the plant licensing bases when the plant is not j generally committed to any specific method to verify seismic l adequacy, but when specific commitments to seismic qualification i requirements exist for equipment connected with the resolution of other specific issues. 13 l l Q-----__-__-___-____________
GW-2., Put [ M. 31 Revision 2 Corrected,6/28/91 For equipment not covered by any specific comm (a) Section 50.59 safety evaluation should be conducted and the L FSAR changed (if appropriate) to reflect the new committront J in the manner suggested for Example 1, above. (l (b) The USI A-46 (GIP) methodology will not supersede seismic qualification requirements imposed or committed to 4 in connection with the resolution of other specific issues (e.g., Regulatory Guide 1.97, TMI Action Item II.F.2, and Individual Plant Examination for External Events). To substitute the USI A-46 (GIP) methodology for specific licensing commitments such as these, licensees must follow Commission regulations, for example,10 C.F.R. 9 50.59 will apply 1n some cases. Examole 3 Revising the plant licensing bases when the plant is generally b ll l committed to_using IEEE 344-1971 to verify the seismic adequacy _ot equipment, but has~no specific commitments to seismic <- g I qualification requirements for equipment connected with the g resolution of other specific issues. Where the plant has a general commitment to IEEE 344-71 and no other specific licensing commitments exist, a Section 50.59 safety evaluation should be performed and documented. In addition, if a change to the FSAR is appropriate, the NRC must be notified pursuant to Section 50.71. ksE[Gd Section 2.3.2 above, the USI A-46 (GIP) methodology is an approved alternative method for satisfying the I ' pertinent equipment seismic requirements of all applicable f I regulations for plants within the scope of USI A-46. , Accordingly, unless there are some unique and unusual l l circumstances applicable to the plant (such as deviations from f the SQUG commitments or implementation guidance of the GIP), a change of commitment from IEEE 344-1971 to the USI A-46 (GIP) h l methodology should not involve an unreviewed safety question fort matters related to verifyin the seismic adequacy _of electrical /, L and mechanicalEmipment. is determination is subject to the l previously-discussed limitation that the USI A-46 (GIP) i methodology applies to equipment seismic adequacy and not to the l overall adequacy of the system in which the equipment is installed. Thus, USI A-46 licensees cre not relieved of the 14 / l
\\ G t P-2., Pod
- N Revision 2 Corrected,2/14/92 requirement to perform and document a Section 50.59 evaluation to f'
determine whether unreviewed safety questions exist. i Examole 4 I Revising the plant licensing bases when the plant is generally committed to using IEEE 344-1971, and in addition, has specific j commitments to other guidance for equipment connected with resolution of specific issues. (a) For equipment not subject to seismic qualification requirements imposed or committed to in connection with the resolution of other specific issues, the plant may modify its commitment to reflect use of the USI A-46 (GIP) i methodology as described for Example 3, above. (b) For equipment subject to seismic qualification requirements imposed or committed to in connection with the I resolution of other specific issues, the commitments to the specific requirements may be revised as described in paragraph (b) of Example 2, above. ([ } Example _} This is a variation of Examples 1-4 that expands the scope of the USI A-46 (GIP) methodology to include equipment outside the scope J of A-46, when the equipment is within the scope of equipment covered by the GIP. (l A change of licensing basis commitment from IEEE 344-1971, as described in above examples, will result in the application of the USI A-46 (GIP) methodology to plant equipment outside the scope of A-46. The scope of application of the USI A-46 (GIP) methodology may be expanded to include this additional mechanical and electrical equipment, provided the licensee also commits to the guidelines presented in Section 2.3.4 of Part I of the GIP, " Future Modifications and New and Replacement Equipment." The Section 50.59 safety evaluation to change the plant licensing bases (as discussed in the above examples) should also consider the effects of expanding the application of the USI A-46 (GIP) methodology from A-46 equipment to the new scope of equipment. However, absent unique and unusual circumstances applicable to the plant, expanding the scope of the A-46 methodology as noted 4 above should not constitute an unreviewed safety question. 15
Re f }} GIP-t, Po rt - L M. s.) ~ Revision 2 m Corrected, 2/14/92 l Subsequent to resolution of USI A-46, if licensees take exception to the GIP criteria and modify those criteria for plant-specific application, this shall be accomplished by modifying the plant licensing bases using the regulatory provisions of 10 C.F.R. 5 50.59. This will be considered a plant-specific modification of a licensing commitment, not a modification of the GIP. 2.3.4 Future Modifications and New and Replacement Eauioment. For any new equipment and replacement of or modifications to equipment having seismic requirements (including equipment not evaluated in response to A-46), licensees shall comply with the plant's licensing bases. Should the licensing bases include use of the USI A-46 (GIP) methodology as an option for verifying seismic adequacy, that methodology will be extended to all mechanical and electrical equipment if and only if the following conditions l are satisfied:8 1. The equipment is reviewed and/or inspected in accordance with the GIP; 2. Equipment changes and modifications are performed in accordance with j the GIP; New or replacement equipment complies with any one of the followingh 3. 5 l a. If it is identical to the equipment originally instalied in the t plant, the criteria and procedures in the GIP apply, 1 i b. If it is not identical to the equipment originally installed in { the plant, the licensee may, on a case-by-case basis establish the equipment's similarity to the installed equipment. The 3 q definition of similarity includes the following elements: (1) 3 excitation, (2) dynamic properties and operability, and (3) dynamic response. After the similarity is established, then the j i criteria and procedures in the GIP apply, or \\, c. If it is not identical to the equipment originally installed in the plant'and the similarity is not established, its seismic ((L adequacy may be verified by conducting a plant and equipment-L 16 Is i
X' ~ ( 6 T-Z, P,,n 24 ~ Revision.2 Corrected,6/28/91 specific evaluation using the approved USI A-46 (GIP) procedures, l or at the licensee's option, application of current seismic qualification criteria or other means acceptable to the staff. j 3 I The A-46 criteria contained herein may be applied in USI A-46 1 J plants to new or replacement equipment not identical to the equipment originally installed in the plant, provided the seismic jj evaluations are performed in a systematic and controlled manner f so as to assure that new or replacement equipment are represented l in the earthqeake experience or generic testing equipment classes and that applicable caveats are met. In particular, each new or replacement item of equipment and part will be reviewed for any / design changes that could adversely affect its seismic capacity \\ from that reflected by the earthquake experience or generic i l testing equipment classes, and these evaluations will be i ( documented. I / The USI A-46 (GIP) methodology is acceptable and sufficient for 3 p verifying the seismic adequacy of commercial grade equipment to be dedicated for safety-related purposes; for other (non-seismic) L j critical characteristics of equipment to be dedicated, licensees j fI are referred to applicable guidance and requirements, such as 09 which include applicable criteria of 10 C.F.R. Part 50, Appen. ) Generic Letter 89-02 (and its supplement) and Generic Letter 89-V d
- 1 ""
4. The GIP is to be maintained in a usable form in the future, with NRC approval of significant changes, in accordance with Section 3.0 of Part I of the GIP. The USI A-46 (GIP) criteria and methodology do not supersede any seismic qualification requirement imposed or committed to in connection with' the resolution of other specific issues (e.g., Regulatory Guide 1.97, Three Mile Island Action Item II.F.2, and Individual Plant Examination for External Events) unless those requirements or commitments are revised according to applicable regulatory requirements. A-46 (GIP) 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) and for new anchorage designs. However, allowable anchorage loads, i.e., factors of safety, currently recommended for new nuclear plants, should be 17' l
C r P z, Putr e, g g Revision 2 Corrected,6/28/91 ] met for newly designed anchorages in modifications, replacements, and new l installations. l The A-46 criteria contained herein may also be applied te new and l replacement cable and conduit raceway systems or parts thereof, in USI A-46 l plants. However, the criteria for evaluation of tanks and heat exchangers, l as defined in Section 7 of Part II of the GIP, are intended only for use on l existing tanks and heat exchangers, not for new installations. When verifying the seismic adequacy of replacement equipment, some flexibility will be allowed in considering the safety function of the equipment. For example, as discussed in Section 6 of Part II, a relay may either be shown to be seismically adequate during an SSE or it may be determined that its function is not necessary for safe shutdown, in which case it is not an essential relay and seismic adequacy need not be %) verified. Similar functional screening is applicable to other parts of replacement equipment. l 2.3.5 Ouality Assurance and Ouality Control. The USI A-46 program for verification of seismic adequacy of equipment as defined by this procedure is outside the scope of commitments made in plant FSARs and Technical Specifications which form the basis of the operating license for the plant; therefore, there is no requirement to perform the USI A-46 program under the nuclear quality assurance and quality control requirements defined for the safety-related equipment in these plants. Instead, the following qualify assurance elements apply to implementation of this procedure for l the USI A-46 program: l 18
m.___.__ SS &c No. 2., Mn[qt Rd.h 1.2.2 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 comp)eted 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 er. tire 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. 1.2.3 Como11ance 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 i 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 electripal equipment _within the scope _ of equipment covered by the GIP...."./The staff recognizes Inat a ncensee may reWse7 (Tis licensing basis in accordance with 10 CFR 50.59 to reflect the [ c acceptability of the USI A-46 (GIP) methodology for verifying the t i seismic adaquacy of electrical and mechanical equipment covered by the l ' GIP The staff's approval of the implementation of the GIP does not i
- funre. relieve the licensees from the requirement to address all aspec viewed safety questions as specified in 10 CFR 50.~59 (for example b
~those plants where_thejihas specified damping values which differ), i from the G_IP.)_(The staff understands tne woro nencefortn to mean, based on SQUE 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 str*f 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. l 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 The may be changed to the GIP seismic riethodology under 10 CFR 50.59. staff has stated, and the SQUG has previously acknowledged, that any previous connitments, 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 I equipment, as described in RG 1.97, the staff agrees that the seismic qualification requirements (RG 1.100, Revision 1) will resolve the US! A-46 requirements for that equipment. The Category 2 and Category 3 equipment as described in RG 1.97 have no specific seismic 7 i.I E
Seut uth,. b&d toluln L Re f. jy - Emergency Procedure Guidelines and Accident Management Guidelines 9 The NRC implemented an Accident Management Program as a result of lessons learned from the accident at Three Mile Island (TMI). The first phase was a short term action to L develop interim procedures for small break loss of coolant accidents (LOCAs) and l . inadequate core cooling. The second phase was development of function-based emergency procedure guidelines (EPGs). The third and final phase was implementation of the EPGs, by individual licensees, into plant specific emergency operating procedures. The Boiling Water Reactor Owner's Group (BWROG) developed EPGs. The NRC approved Revision 4 of the EPGs in NEDO-31331, March 1987 (Reference 1) in a safety evaluation report (SER) dated September 12,1988 (Reference 2), and requested licensees j to revise their emergency operating procedures to reflect the guidance in the EPGs, Revision 4, as early as practical. Consistent with the SER, which stated that licensees wishing to use Revision 4 of the EPGs should assure the implementation did not impact their licensing bases, licensees implemented the revision under the provisions of 10 CFR 50.59. Ongoing work in the accident management area resulted in the more recently developed Accident Management Guidelines (AMGs) developed by the BWROG The AMGs will be used to add severe accident management methodologies to the EPGs. In a letter from Mr. Ashok C. Thadani of the NRC to Mr. S. Labruna of the BWROG, dated May 27,1994 (Reference 3), the NRC stated that the "AMGs and any associated EPG changes may be implemented in accordance with 10 CFR 50.59." The letter also indicated the NRC does not intend to issue a safety evaluation report on the AMGs. The incorporation of the EPGs and AMGs into plant specific EOPs implements the overall accident management methodologies whien were developed as a result of significant effort h within the industry and the NRC. The NRC considers these methodologies to be a f significant improvement in the safety of nuclear plants and that implementation of these l methodologies will enhance accident prevention and mitigation, as well as reduce potential consequences. In the Policy Statement on Severe Reactor Accidents Regarding Future y Design and Existing Plants, dated August 8,1985, the NRC did not attempt to quantify / each of the numerous changes to the equipment, procedures, analyses and reports 1 associated with the TMI Action Plan. However,it came to the conclusion that the / cumulative effect of all these changes provided a safety improvement: k- "There were 132 different types of action items approved in the [TMI) Action } - Plan (an average of 90 actions per plant). Of this total,39 actions involved 1 equipment backfit items,31 actions involved procedural changes, and 62 actions required analyses and reperts. It is impractical to quantify all of the safety improvemem s obtained by these many c.hanges. Nevertheless, the cumulative effect is undoubtedly a significant improvement in safety." - _ - _ - - - - - - - - -, - - -, - - - - - -. _ - - _ - - - - - - - - - - - - - -. - - - - _ ~. - - - - - - - - -
$ W(> l.a Yer ba$rb toln/A y b(lesun L Esf, /]/ 1 Both the EPGs and the AMGs consist of accident management methodologies which may h be considered to contain less conservative elements than those in plants' licensing bases. I For instance, the EPGs direct the operators to inhibit the automatic actuation of the safety relief valves (SRVs)in the reactor automatic depressurization system (ADS). This action is 1 ( not consistent with the design basis automatic safety function of the ADS. Additionally, the j AMGs direct the operators to vent the containment, even in the event of known core damage. These elements, when taken alone, could be perceived as decreasing the margin (of safety when compared to the plants' original licensing bases y The NRC accepted implementation of the actions requested as a result of the Accident Management Program based upon the fact that implementation of the methodologies as a whole represents an increase in the safety margin of the plants.
References:
1. BWR Owners' Group - Emergency Procedure Guidelines, Revision 4, NEDO-31331, March 1987. 2. Safety Evaluation of BWR Owners' Grot.p - Emergency Procedure Guideline's, Revision 4, NEDO-31331, March 1987, dated September 12,1988 3. NRC letter from Mr. Ashok C.Thadani of the NRC to Mr. S. Labruna of the BWROG, dated May 27,1994. 4. Policy Statement on Severe Reactor Accidents Regarding Future Design and Existing Plants dated August 8,1985. __-
f - \\ 5 G.u G L< de r kind olufn Ref}ll Inservice Inspection of Containment Tendons - _ - _ ~ In 1976 the NRC issued Revision 2 of Regulatory Guide (RG) 1.35 to define methods for performing inservice inspection of ungrouted tendons in prestressed concrete containment i structures. This guide defines how samples are to be selected and what types of inspections I should be performed. As a result oflessons learned from implementing this guidance and a, more complete understanding of potential failure modes, Revision 3 to this regulatory ~- (, guide was issued in 1990.f Some of the changes made in Revision 3 of RG 1.35 clarify certain testing criteria and reporting guidance. However, a number of changes were__made which relaxed requirements in tlye previous revisionfor example in Kevisioli"2, all tendons in the sample are required o be detensi5ned to look for evidence of broken or damaged wires or strands; Revision 3 o ly requires one randomly selecte roup to be detensionedf Another example of relaxed requirements relates to_the measured tendon prestressi_nn force during monitoring tests./5Eevision 2 oTRF1.35 has a stringent requirenient that the measured prestressing force must be within limits prescribed for the time of the test; Revision 3 relaxes this requirement so that a reportable condition is considered to exist only ( when the measured prestressing force of any tendon lies below 90% of the prescribed lower limit and the prestressing force in two adjacent tendons lies below 95% of the prescribed lower limit, or the average of all measured tendon forces for each group is found to be less than the minimum required prestress level for the groupf ~~ Although certain elements of Revision 3 of RG 1.35, taken by themselves are less conservative than comparable elements in Revision 2, there are additional requirements in Revision 3 which aodress weak areas and potential failure modes not recognized in Revision 2. Fur example, Revision 3 requires an additional visual irtspection of the bottom grease caps of all vertical tendons to detect signs of grease leakage or grease cap deformations. Also, Revision 3 requires additional testing if an adverse trend is found where loss of prestressing force is greater than expected during consecutive surveillance l inspections. Revision 3 also requires examination of sheathing filler grease to determine water content, reserve alkalinity, and concentration of water-soluble chlorides, nitrates, and ' sulfides. J The change from Revision 2 to Revision 3 of RG 1.35 illustrates that the NRC staff accepts an alternate evaluation methodology even though elements may be less conservative than de methodology being replaced, provided the potential failure mechanisms are better i ' understood and compensating effects are evaluated, even if these tradeoffs cannot be quantified. A regulatory analysis (Reference 1) sponsored by the NRC on a draft version of Revision 3 of RG 1.35 concluded that:
\\ SQVG Leder kfel m/ajn bliswre 3 leef.Lu "Although changes in the guide were determined to produce an unquantifiable change in risk, it is anticipated that a reduction in the l number of tendons to be detensioned and the increased requirements for visual inspection, grease impurity levels, and prestress monitoring of the second containment (at a two unit site) will have a positive impact on safety and, thus, will lower the risk. Containment availability should, therefore, be enhanced." l eference 1, page 46) J There are several parallels between changing from Revision 2 to Revision 3 of RG 1.35 and changing from the seismic qualification requirements contained in the licensing basis of A-46 plants to the GIP as supplemented by SSER #2. Just as there were lessons learned ,J between issuance of Revision 2 and Revision 3 of RG 1.35, significant lessons were learned and incorporated into the GIP, particularly as a result of evaluating the performance of industrial grade equipment subjected to real earthquakes with significant ground motion. ? The experience data collected from numerous earthquakes at many sites, containing / hundreds ofitems of equipments,were used to develop a balanced set of seismic criteria j and guidelines and placed in the GIP. The GIP methodology includes requirements to / review certain features not typically included in the licensing basis of A-46 plants. For l example, the GIP checks for known seismic vulnerabilities (caveats) in various types of equipment, a check is made for seismic interaction with nearby structures (including h sufficient slack in service lines), and as-installed checks are made of anchorage adequacy h I (including visualinspection of all anchors, torque tests of expansion anchors, edge distance h and spacing checks, etc.). These additional GIF requirements along with certain other h (h relaxed requirements are based on a more thorough understanding of the effects which strong ground motion earthquakes have on industrial equipment. The NRC concurred in this overall approach and accepted use of the GIP as an alternative methodology in ( SSER #2. References (1) NUREG/CR-4712, Regulatory Analysis of Regulatory Guide 1.35 (Revision 3, Draft 2) " Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments," by Oak Ridge National I_aboratory. 1 l
SadC. Litte haies rolaba p Eef,3/ BWR Mark I Containment Long Term Criteria '[ Boiling Water Reactors (BWRs) with Mark I containments which were built dur 1960s and early 1970s used versions of the ASME Boiler and Pressure Vessel Coft which were in effect at the time, ne 1%5 and 1968 editions of the Code, Section III designated metal containment structures as Class "B" vessels and allowed use of design by analysis 1 rules based on the Basic Stress Intensity Limits.' Dese early editions of the Code did not j provide higher allowables for infrequent " Emergency" or " Faulted" conditions for these (ctypes of vessels as permitted in later editions.f . When it was discovered in the mid-1970s that certain suppression pool hydrodynamic loads may not have been considered in the original design of the Mark I containments, the NRC identified an " Unresolved Safety Issue"(Task Action Plan A-7). An industry group formed j (Mark I Owners Group) and conducted experimental and analytical programs which defined these new hydrodynamic loads in a I. cad Definition Report (LDR) (Reference 1). l Structural assessment techniques were defined by the ownersfroup in a Plant Unique Analysis Application Guide (PUAAG)(Reference 2).fThe3urnmer 1977 edition of the fASME Boiler ana Pressure Vessel Code, Section III was used as the basis for the structural ! assessment techniques. This edition of the ASME Code allows less restrictive stress Qntensity limits for certain infrequent loading conditions.8f Even though these newly discovered loads should have been considered in the original design of the Mark I containments, the NRC in their Safety Analysis Report (SER), NUREG-0661 (Reference 3), allowed licensees to use the PUAAG (which is based on the 1977 edition of the ASME Code) as the basis for evaluating the new loads defined in the LDR. This was allowed even though the NRC expected,"... all other loading conditions and structurn analysis techniques (e.g., dead load and seismic loads) [to) be in accordance with the plant's Final Safety Analysis Report (FSAR)."(Page A-4 of Reference 3) ' Basic Stress Intensity Limits in Section III of the ASME Boiler and Pressure Vessel Code include limits such as 1.0 S,,, for general primary membrane stress and 1.5 S,, for primary membrane plus bending stress. 'The 1.evel C Service Limits in the 1977 edition of the ASME Boiler and Pressure Vessel Code allow higher stress intensity limits than Basic Stress Intensity Limits. For example, the greater of 1.0 S, or 1.2 S. is allowed for general primary membrane - stress and the greater of 1.5 S, or 1.8 S, is allowed for primary membrane plus bending stress.
saac u ed sa A to Eulosne & geny Thi., example of using alternate acceptance criteria for the Mark I Containment Program illustrates that the NRC accepts more realistic,less conservative analysis techniques, particularly when the loadings are more accurately defined and areas of uncertainty are addressed. I The BWR Mark I Program is similar to tha USI A-46 program in some respects. The Mark I Program was an industry-led activity in which alternate loads and evaluation methods were defined and accepted by the NRC in a safety evaluation report. Even though c there were elements of the generic program which, taken by themselves, could be viewed as ] being less conservative than those in the plant licensing basis, many licensees adopted, with NRC knowledge, the Mark I program load definitions and analysis techniques as a part of l their plant licensing basis using the 50.59 process on the basis that the overall safety of the plant was enhanced. References (1) General Electric Topical Report NEDO-21888," Mark I Containment Program,I. cad Definition Report," Revision 0, December 1978. (2) General Electric Topical Report NEDO 24583-1," Mark I Containment Program, Structural Acceptance Criteria, Plant-Unique Analysis Applications Guide, Task 3.1.3," October 1979. (3) NRC NUREG-0661," Safety Evaluation Report, Mark I Containment Long-Term Program, Resolution of Generic Technical Activity A 7," July 1980.
$Q.UG Le $er Yest CILLf47 Itef,2] ASME Code Case N-411: Damping for Piping Systems Seismic dynamic analyses are used to show that structures, systems, and components remain functional during and after earthquakes. Viscous damping is used in these analyses to account for the energy dissipated during dynamic motion. ( The original licensing bases of early nuclear plants typically specify damping values of 1/2% ' to 2% for the dynamic analysis of piping systems. The NRC accepts higher values of L damping provided the licensee considers other characteristics of the piping system together with analysis techniques currently in use for new plants. j- [ For example, Regulatory Guide (RG) 1.84 allows ASME Code Case N-411 to be used with L damping values which range from 2% to 5%, depending upon the natural frequency of the j piping system. These higher,less conservative damping values are accepted by the NRC L provided current analysis techniques are also used (e.g., RG 1.60 spectra shape, RG 1.122 h spectra broadening or Code Case N-397 peak shifting, RG 1.92 modal and directional ' combinations, etc.). Further, N-411 damping is not allowed (1) for piping where stress I corrosion cracking has occurred, (2) for time-history analysis techniques, or (3) for piping l t ystems with energy absorbersg - ~ s This ex::mple illustrates that the NRC staff accepts more realistic,less conservative analysis techniques provided additional areas of uncertainty are addressed and the methods are applied in their entirety. l 1
GS 01 tuo.*2, W n l9 N af. y qualification provisions. Therefore, if that equipment is used as part f' 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 A guidance and requirements as GL 89-02, GL 89-09, and GL 91-05, which V include applicable criteria of 10 CFR Part 50, Appendix B. 4 The staff normally would require that new or replacement equipment be cualified in accordance with plant-specific licensing commitments or { current criteria (e.g., 10 CFR 50.49) unless the licensee can justify l l 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 i 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 i 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 I for Resolution of Unresolved Safety Issue A-46, ' Seismic Qualification of Equipment in Operating Plants'," was that the criteria and procedures j described herein are determined to be an acceptable evaluation method for verifyina 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 :urrent seismic qualification requirements, whereas the seismic adequacy of all other safe shutdown equipment (which will 8
f SS C72-t)o. 2, Gatf9 z Ed. y 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 verifyin.g the seismic adequacy of new n equipment in US! 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 l and approval, a procedure for evaluating the acceptability of new data and a l procedure for updating and revising GIP-2 and subsequent revisions, based on new information including the lessons learned during the US! A-46 plant wal kdowns. 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.I introduction Section 1 of Part !! 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 e screening verification end walkdown outlier identification and resolution a relay functionality review tanks and heat exchangers review cable and conduit raceway review documentation 11.2 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 The engineers will develop the list of equipment required for safe shutdown. systems engineer should be a degree engineer, or equivalent, and should have had extensive experience with, and broad understanding of, the systemt, equipment, and procedures of the plant. The seismic capability engineers will conduct the walkdowns and assess the seismic adequacy of safe-shutdown 9 I
NOC La kW h o ( M G [ 9 '7 [2e.f,g &cloSurt by the GlP, The staffs approval of the implementation of the GlP does not relieve the licensees from the requirement to address all asoects [ emphasis added] of unreviewed safety questions as specified in 10 CFR 50.59." (The staff notes that SQUG's October 6,1997, letter, did not address this portion of the staff SER). As stated during the October 30,1997 meeting, and as recognized on page 4 in the staffs SER, "the implementation of the GIP-2 approach for USl A-46 plants provides safety enhancement, in cedain ascects, [ emphasis added) beyond the originallicensing basis." The primary safety enhancements realized from implementing the USl A-46 resolution via the application of GlP-2, are attributed to the walkdown of equipment and associated reso %n of identified discrepancies in the safe shutdown path. There are design relateo a spects in the GlP-2, however, that may not be as conservative as specific licensing basis req 2ements delineated for the qu'alification of new equipment in USl A-46 plants.) Ea'ch licensee is free to conduct an 7 analysis of the applicability of the UW to the ucensing oasis in the FSAR using the provisions of 10 CFR 50.59, but the analysis cannot simply conclude that there is no USO merely on the basis l that the GlP methodology, taken as a whole, represents an improvement. Since each A-46 j plant FSAR may contain specific or unique commitments, each licensee must review the various / aspects of the GIP methodology against the various related aspects of the plant licensing basis (and__ determine, after reviewing the specific aspects, if the adoption of the GIP involves an USQ. If the licensee concludes that a US6IsTnvolved, adoption of the GIP into the licermiB basis ~ 9 would require the licensee to seek an amendment under the provisions of 10 CCR 50.90. Consistent with the staffs May 22,1992, SER, the staff intends to reach closure on USl A-46 implementation at a facility prior to approving an amendment to incorporate the GIP into the f licensing basis (should one be proposed) for that facility. ) As an example of the above staff position, the staff considered the situation where equipment damping is specified in a plant's FSAR and the licensee seeks to make an evaluation under the provisions of 10 CFR 50.59. If the FSAR contains specific equipment damping values that are less than those permitted by the GIP, the licensee must review the aspects affected by the change and make its own determination regarding whether or not an USQ exists at its facility. An analogous situation (discussed with SQUG during the October 30,1997, meeting) involved the application of seismic demand criteria. If a licensee determines that a license amendment is the viable means to approach the GIP-2 l incorporation into the licensing basis, an application should be submitted requesting the desired change. To facilitate the staffs evaluation, the application should contain: 1) identification of areas in the GlP (e.g., criteria, procedures, methodologies) that the licensee proposes to incorporate in the FSAR: 2) aspects of the GIP where the proposed change is less conservative than the FSAR criteria specified for that particular aspect (e.g., non-safety vs. safety equipment);
- 3) identification of the plant's components and equipment to which the gip will be applicable; and 4) justification to support the adequacy of the proposed change. Referencing appropriate portions of the licensee's USl A-46 submittal and the staffs plant-specific SER is suggested.
l The staffs review of amendment requests should be facilitated by the review effort already I expended on the licensee's USl A-46 program. l In summary, neither the staff nor SQUG can reach a generic conclusion regarding the appropriate means to incorporate the GIP-2 into a facility's licensing basis. The determination is ) l based primarily on the specifics in the licensing basis for the particular facility. We believe that a i constnde dialogue between the staff and the lead plants proposing to amend their licenses 1 y. ~
E o 3 NUCLEAR REGIUTORY COMMISSION ~ )f }, t. k nswmstow.n.c. - 5,owf %,~.
- Dune 20, 1997 Nfic. Le.$r Nest Giro(91
.,,.L Mr. H. L. Sumner, Jr. ~ Vice President Southern Nuclear Operating g g* 5/- f Company, Inc. P. O. Box 1295 Simingham, Alabama 35201-1295
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING INCORPORATION OF THE GENERIC IMPLEMENTATION PROCEDURE IN THE FINAL SAFETY ANALYSIS REPORT (FSAR) UNDER THE PROVISIONS OF 10 CFR 50.59 - EDWIN I.-HATCH NUCLEAR l PLANT, UNITS 1 AND 2 (TAC N05. M69453 AND M69452)
Dear Mr. Sumner:
On September 16, 1992, Georgia Power Company (GPC) submitted the initial res;,voss :.o Supplement He. I to tieneric Letter (GL) 87-02, " Verification of Seis=f c Ade:cacy of Mechanical and Electrical Ecuipment in Operating Reactors, Unresolved Safety Issue (USI) A-46,' dated May 22, 1992, regarding USI A-46 - f:r the Edwin I. Hat:h Nuclear Plant, Units I and 2 (Plant Hatch). In that subst:tal, GPC indicated its intent to incorporate provisions of the Generic Implementation Procedure (GIP) into the F5AR, employing the provisions of 30 CFR 50.59 u:en re:eipt of a final plant-specific Safety Evaluation (SE) resolving USI A-46. - The GIF was develeped by seismic Qualification Utility Gr:up (SOUG).specifically for use in resolving USI A-46. The staff approved the use of the GIP in the SE dated May 22, M92, issued with Supplement I to O-GL 37-02. The staff's SI stated that a licensee may revise its licensing basis in a:::rdance with 10 CFR 50.59 te reflec: the ac:ep: ability of the U51 A-s6 (GI?) methodology f:r verifyin; the seismic ace:uacy cf ele::rical and mechanical e:uipment covered by the GIP; however, the staff assumed that such int:rp:ra:i:n weuld n t lead ::.a reduction in the licensees' licensing basis margins. In a letter :: the NRC on July 31, 1996, GPC informed the staff that i had : hanged the plant licensing basis by incorporating the USI A-46 GIP me:h:d: logy into tha'F5AR in Revision 14C. In a re:uest for adcitional inf:r=atien cated Jan,uary 30, 1997, relating to GPC's July 31 and August 23, 2955, su:mittals on the U51 A-46 im;1ementation a: Plant Hat:h, the staff re:ueste: GPC :: provide c:mplete documentation asso:ia ed witn the 50.59 evaluz i:n. G?C submitted the re;uested documentation in its resp:nse of A:ril 25,1997. Based on a review of GPC's April 25, 1997, response, the staff identified a p::ential violation involving the subject 50.59 evaluation. The F5AR revision veuld per=it, as an alternative, the use of the GIP procedures for the seismic cualifi:stion of all me:hanical and electrical equipment in the plant. The SIP is only ac:eptable for resolution of USI A-46 seismic adequacy issues and may be used for verifying the seismic adecuacy of new or replacement equipment as stated in the staff's SE of May 22, 1992, when 50.59 criteria are met. Th'e staff believes certain Srovisions of the GIP may not be consistent with Plant Hat:h licensing basis. Among the provisions in the GIP that might not be c:nsistent-with the F5AR' licensing basis criteria and which were not r: dressed % GPC's 50.59 evaluation, is a provision that per=its the use of ' seismic capacity spectrum equal to 1.5 times the SQUG Bounding spe:trum for
~ [ Ngc t e b Md c/eh, H. L sumner, Jr. f Ref.k] ~F the seismic evaluation of equipment located on floors within 40 feet above the effective grade level. For Plant Hatch, the SQUG capacity spectrum is c:nsiderably lower than the FSAR in-structure response spectra (IRS) for some floors in the reactor building at certain frequency ranges. GPC in its April 25,1997, response, stated that in performing the A-46 evaluation, it has employed a demand spectrum equal to one-half the Plant Hatch seismic margin earthquake IRS, which was specifically approved by the staff for the erelicit purpose of USI A-46 resolution at the facility. The spectral accelerations in the IRS that are developed from.5 times the seismic margin earthquake spectrum are lower than the capacity spectrum that is based on l 1.5 times the SQUG Bounding spectrum for floors within 40 feet above the grade level. In applying the GIP (in lieu of previously approved F5AR criteria) as a licensing basis criteria for the qualification of mechanical and eltetrical e;uipment, GPC c:uld usa either 1.5 times the plant safe shutdown earthquake ground spectrum or the seismic margin earthquake IRS times.5 as the required demand spectra for equipment within 40 feet above the grade level. At some floor elevations, either of these spectra may underestimate the seismic demand when c:=;ared to the F5AR response spectr,a (licensing basis) at certain frecuency ranges. The seismic qualification of mechanical and electrical equipment to the GIF provision, as discussed above, could potentially result in an underestimation of the equipment seismic demand. This can result in the eventual reduction to the licensing basis margins for the compenents or systems affected by such changes to the facility and, consequently, would not b O e c:nsistent with the limitations of 10 CFR 50.59. l As stated absve, GPC's 50.55 evaivatien to incorporate the GIF in the FSAR cid not address ether potential areas of c:nflict between the licensing basis . criteria and the GIP document. These areas incluce (but may not be limited tc), in adcition to the seismic design spectra issue discussed above, variation in dampingtrecuirements for e:;uipment, specific re;uirements for cualificatien by testing er analysis for certain e:;ui; ment, and the use of the 'ruie-of-ths-box" endorsed in the Gi?. Furthermore, the staff has aiso identified that GPC has ceviated from its original c:mmitment in the September 15, 1992, respense to G1. 87-02 on USI A-46 resolution. In that respense, GPC stated that Plant Hatch intended te affect an TSAR change by. incorporating the GIP-2 procedures via a 50.59 evaluation upon receipt of a Nai ;iani-specific SE resolving USI A-46 (a condition affirmed in the staff's May 22,1932, SE). As of the date of this letter, the staff has not issued its final SE on the A-45 implementation at Plant Hatch. In the meantime, GPC has already incorporated the GIP into the FSAR Revision 14C for Units 1 and 2. i Based on the above discussion, the staff concludes that GPC's action of fl997 changing the FSAR as discussed in the letters of July 31, 1996, and April 25 tp_ the NRC. involves a potential 50.59 violatientas. well as a deviation from GPC's c:mitment in its 5eptemoer 26, 1992, letter to the staff. The staff is unaware whether the GIP procedure has been employed by GPC in l ( a
- 'M. L. Sumner, Jr. *:
NflQ Qf{s, & hed lol7W9 ') Tr f. b) 7 ),. actually making a change to the facility outside the scope of USI A-46. Therefore., we request that you provide your response to resolve the above staff concerns within 60 days from the date of this letter. If you have any questions, please contact me at (301) 415-1458. Sincerely, Ab, t l Ngoc B. Le, Project Manager Project Directorate II-2 Division of Reactor Projects - I/II l Office of Nuclear Reactor Regulation l Docket Nes. 50-321 and 50-366 1 cc: See ner: page 1 0 l L t ( i l I i
h)RC Ilaa/mi & 9/4(9 7 owT U I" ta s<L Jb.SQuc gq 3 FRAMEWORK FOR USE OF GIP-2 TO DEMONSTRATE SEISMIC ADEQUACY OF NEW AND REPLACEMENT EQUIPMENTIN USI A-46 PLANTS PREREQUISITE Complete resolution of IJSI A-46 at the facility and issuance of staff closure SE AREAS TO ADDRESS A. Plants whose current licensing basis do not include specific commitments to seismic qualification approaches, methods or parameters used in conventional seismic analysis methodologies I e Licensees may use 50.59 to incorporate GIP-2 in the FSAR for equipment classes covered by GIF-2 Discuss GlP-2 applicability to overall population of plant e components and equipment Specify seismic qualification approach (es) for new and e replacment equipment (NARE) that are neither identical nor similar to equipment classes covered by GIP-2 __m B. Plants whose current licensing bases include specific commitments to seismic qualification approaches and methods or parameters used in i conventional seismic analysis methodologies j l identify the criteria and procedures (including parameters) e specified in GIP-2 that deviate from, and are less conservative l .l l than, seismic design attributes specified in current licensing l basis (e.g., plant-specific in-structure spectra, equipment damping values, etc) l l Identify the criteria, procedures, parameters, etc., in GIP-2 that t e (t deviate from, arid are less conservative than, licensing l I commitments (e.g., IEEE-344, industry Standards, etc.) related to equipment-specific qualification aspects. Particular areas to l [ be addressed include. l 1
l N(2C. Hoa6,t uY 9/ +M 7 Nfia; w/Sa v6 flef'. 7_l l 1 1) Use of non-safety-related equipment versus safety-related ) equipment or Class 1E equipment; ( 2) Use of the " Rule of Box" versus component or device l testing or analysis; } / 3) Use of" satisfying the intent of equipment caveat" versus g actual equipment dynamic characteristics or similarity; and / 4) Fulfillment of requirements on environmental qualification ) ll of equipment (i.e., applicable requirements such as 10 CFR 50.49, NUREG-0588, IEEE 323-1974 or 1971). Submit a license amendment to address identified deviations i and provide justification for future use of GIP-2 with (or w l a plant-specific revision, to demonstrate seismic adequacy of { NARE that are identical or similar to equipment classes covered / g by GIP-2 within (or outside) A-46 scope fa Discuss GIP-2 applicability to overall population of plant lg [ components and equipment f e Specify seismic qualification approach (es) for NARE that are lm !E neither identical nor similar to equipment classes covered by GIP-2 ~ l STAFF REVIEW / APPROVAL PROCESS k Based on supplemental staff SE on SQUG's GIP-2 and plant-e specific A-46 resolution SE E Criteria in staff draft guidelines presented in August 14,1997, e g meeting with SQUG Thorough account and adequate reconciliation of deviations e ,k between currentlicensing basis and proposed FSAR change 1 g Expedient review depending on volume of amendments e
N0( Coe m M Ms A hoh W' %kt Power Oconee o, Three License Conditions: Q A y u 4,gy 4/qg 1. Complete the necessary SEWS forms for equipment determined to be seismically adequate using GIP-2. 2. Provide more detailed comparisons of the critical characteristics (i.e., manufacturer, model, materials, physical properties, dimensions, dynamic characteristics, etc.) for the new equipment with data from testing or recorded earthquakes in the experience databasy 3. Add the equipment in the ECCW system to the USI A-46 SSEL and include its evaluation in a. revision to the USl A 46 submittal. I I l
Da M # I 9d. LO.l 4.2.1 10 CFR 50.59 Change Process A licensee may change his licensing basis under 50.59 provided (1) the probability and consequences of an accident are not increased, (2) a new type of accident is not introduced, and (3) there is no reduction in seismic margin. The probability and consequences of an accident and the potential for a new type accident are not changed by the addition of the GIP to the licensing basis provided the GIP provides at least as much assurance of equipment seismic adequacy as the existing licensing basis method (s). Further, applying the GIP as a whole does not reduce the overall seismic margin of the plant. Specifically, for A-46 plants, the GIP method, taken in its entirety,2 is considered to provide improved seismic margin for equipment and raceways in its scope for the following reasons: = = 1. The GIP is a new and different methodology, based on use of earthquake and test experience and supplemented by analysis. It is in effect a "new model"3. As such, consistent with accepted industry and NRC practice, the seismic margin provided must be evaluated on the overall program level. Comparison of individual elements of this new methodology with elements of the existing licensing basis is not appropriate or valid. This is also consistent with the explicit requirement of SSER No. 2,2 that the GIP be applied in its entirety. 2. The GIP method was throughly reviewed by the SOUG licensees, an independent panel of seismic experts, NRC consultants and NRC staff, and was accepted in SSER No. 2. 3. The GIP meets the applicable regulations. After recognizing differences in the GIP method compared to current seismic qualification methods, the NRC states in SSER No. 2, pages 4 and 5, that the GIP l ... approach for USI A-46 plants provides safety enhancements,in certain aspects, beyond the original licensing basis. Therefore, GIP 2 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 2SSER No. 2.(Reference 2), pg. 6 1 'NUREG-1211 (Reference 4), pg.16 I
1 I 5 QVC Leiler %fd to/s {97 t ( M el l l 4. Adontion of the GIP as a Licensing Basis Method for NARE Given that it is acceptable,in fact desirable, to use the GIP method to maintain the level of seismic adequacy achieved through the USI A 46 program by also applying i the GIP to new and replacement equipment, all that remains is to determine the l appropriate mechanism to adopt the GIP for this purpose. Because the A-46 seismic review was performed pursuant to a 50.54(f) information request, implementation of j l the GIP to resolve USI A-46 is a " snapshot in time" and does not change any plant's seismic licensing basis. Therefore, for plants whose current seismic licensing basis l requires use of specific seismic qualification methods (e.g., IEEE 344-1971), a formal change to the licensing basis is required to adopt the GIP as an acceptable alternative method for equipment and raceways within the scope of the GIP. For plants which j have only generallicensing basis commitments (e.g., to design the plant to withstand a ] i design basis earthquake), the GIP method is likely consistent with the existing licensing basis, such that no licensing basis change is mandatory. Guidance and options for how to adopt the GIP for NARE in these two cases are given below. -~ r, c ~ ~. -._ 4.1 A-46 Plants with General. Non Specific Utensing Bases 1 General, non-specific licensing bases are those which refer to no specific ) i standard, method or procedure for demonstration of seismic adequacy of i equipment and electrical raceways. Examples include general statements that l such equipment "shall be designed to withstand a design basis earthquake," 4 "shall be shown to be seismically adequate by analysis or test"(without ) specificity), and the like. For those cases, the GIP methodology may be ( l considered an acceptable engineering method for demonstrating seismic ( adequacy that is consistent with the licensing basis. Thus, no formal change to I the licensing basis would be mandatory. Should the utility choose to incorporate the GIP method as an acceptable method in the plant FSAR, a 50.59 safety evaluation based on the guidance of 4.2, below, would not involve an Unreviewed Safety Question (USO). ~ - - - - - 4.2 A-46 Plants with Specific Licensing Bases Plants with licensing bases which reference specific standards or processes for determining seismic adequacy (e.g., IEEE 344-71) should change their seismic l licensing bases to adopt the GIP as an acceptable licensing basis method. Two l approaches are provided in the regulations to do this. I SOU( Le t{ac Nfd wh/97 Lf:pu requirements of General Design Criterion 2 and the purpose of the NRC regulations relevant to equipment seismic adequacy including 10 CFR Part 100." 4. The NRC and SQUG agree that implementation of the GIP method as a complete package enhances the seismic safety of the A-46 plants, as intended.' Based on the above,it is clear that the use of the GIP methodology does not re6ce seismic margins relative to the existing licensing bases; therefore the 50.59 safety evaluation should not involve a USQ.8 4.2.2 10 CFR 66 50.90. 50.91. and 50.92. License Amendment Process A second way to change a plant's licensing basis is by license amendment in accordance with Sections 50.90,50.91, and 50.92 of the regulations. While this method would not be required unless special licensing basis commitments preclude use of the 50.59 process,it is an option for all A-46 plants. license amendment must be submitted to the NRC for approval and, to be successful, must show that the proposed change (i.e., adoption of the GIP methodology) provides idequate protection of the public health and safety. A comparison of the thange with the existing licensing basis similar. j to the 50.59 evaluations is also required, but only to determine if any f requested public hearing is held before or after approval of the l amendment. No comparison with the existing licensing basis is required ( order to establish that the standard of adequate protection of the publicis } l met. 1 Because it has been determined by the NRC that the GIP method, taken it is entirety, meets the applicable requirements of GDC-2 and 10 CFR 100, Appendix A,it also meets the standard of adequate protection of the health and safety of the public by definition. Thus, the test for a successful licensing amendment is met solely by the fact that the GIP meets the plicable regulations. l 'During the meeting held between SQUG representatives and the NRC staff on August 14,1997, Messrs. Wessman and Bagchi agreed that implementation of the USI A-46 program, including resolution of outliers, enhances the seismic safety of the A-46 plants. L 8A similar conclusion is reached in GIP-2 (Reference 1), Part I, Section 2.3.3, Example 3, pg.14; and SSER No. 2 did not challenge that conclusion. L
s m u,, w ,o4&/n I Hardened WetwellVent (GL 8916) NRC Generic I.ctter (GL) 89-16 (Reference 1) asks licensees with BWRs to install a vent path from the containment wetwell to the stack to avoid the potential for over pressurization and rupture of the containment during a postulated severe accident condition. This request is based on the results of two methodologies: the probabilistic risk assessment methodology (implemented by the industry per GL 88-20) and the accident management methodology (voluntary industry implementation by owner's group initiative). Application of these two methodologies indicated that the addition of this containment vent capability (1) can reduce 'he likelihood of core melt from certain accident sequences (i.e., reduces core damage fr equency) and (2) meets the objectives of the emergency procedure ' elines which require venting the containment under specific accident scenario us, when considering the resuhs of the methodologies as a whole, implementation of this modification increases the safety margin of the plants. SECY 89-017 (Reference 2) which v presents the Mark I Containment Performance Improvement Program, undertaken for i closure of severe accident issues, states: "These improvements, although not representing large changes to the plants, form an integrated set which, when fully implemented, will substantially i enhance the safety of Mark I plants by enhancing defense in depth, including improvement to containment performance." __ 2 - _.. - - m her, when co'nsiilering one element of this recommended plant change (i.e., the s installation of a hardened wetwell vent compared to the plants' original licensing bases t (e.g., deterministic methods with event-related emergency procedures), the conclusion that L the safety margin is increased is not directly apparent. This modification installs a direct ( vent path from the primary containment to the atmosphere, bypassing the secondary L containment filtering systems. The NRC recognized (in Reference 3) that there are L potential drawbacks ("downside") to installing containment vents. Therefore, without ' considering the overall results of the probabilistic and accident management methodologies,it could be said the margin of safety is decreased with installation of the vent. Nevertheless, the NRC accepted implementation of the actions requested in GL 89-16 based upon the fact that implementation of the methodologies, taken as a whole, represents an increase in the safety margin of the plants. In fact, the staff"strongly encouraged licensees to implement requisite design changes... under the provisions of 10 CFR 50.59." (Reference 1) I l L-_-___-_-____-__
SQuc Lt he bda nhz(9, 6m elosure j (2e{g 1. Generic letter 89-16, Installation of a Hardened Wetwell Vent, dated September 1,1989. 2. SECY 89-017 dated January 1989. 3. NRC (A. U Thadani) letter to BWR Owners' Group (D. Grace) dated September 12, 1988, Safety Evaluation of "BWR Owners' Group - Emergency Procedure Guidelines, Revision 4," NEDO-31331, March 1987. i I [ ( 2- _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _}}