3F1298-01, Forwards Response to NRC 980410 RAI Re Audit Rept of USI A-46,seismic Implementation & Subsequent Evaluations of Related Issues at Crystal River,Unit 3
| ML20198N304 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/31/1998 |
| From: | Roderick D FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR 3F1298-01, 3F1298-1, TAC-M69440, NUDOCS 9901060089 | |
| Download: ML20198N304 (27) | |
Text
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December 31,1998 3F12934)1 j U.S. Nuclear Regulatory Commission Attn: Document Control Desk {
Washington, DC 20555-0001 i
Subject:
Response to NRC Request for Additional Information Regarding Unresolved Safety Issue (USI) A-46 (TAC No. M69440)
)
References:
- 1. NRC to FPC letter, 3N0498-07, dated April 10,1998, " Audit Report of Unresolved Safety Issue A-46, Seismic Implementation and Subsequent Evaluations of Related Issues at Crystal River Unit 3 (TAC No. M69440)*
- 2. FPC to NRC letter 3F1297-24, dated December 16,1997, " Supplemental Response for the Resolution of Unresolved Safety Issue A-46 (TAC No. M69440)"
- 3. FPC to NRC letter 3F0.'98-16, dated March 30,1998, " Request for l Additional Information Regarding Summary Report on the Verification of Seismic Adequacy of Mechanical and Electrical Equipment dated December 31,1995"
Dear Sir:
Reference 1 forwarded the results of the NRC Staff audit of the Unresolved Safety Issue (USI) /
l A-46 program at Crystal River Unit 3 (CR-3), and requested additional information.
I Enclosure 1 of Reference 1 identified numerous concerns regarding the USI A-46 program that remain to be resolved. Enclosure 2 of Reference 1 identified specific concerns regarding cable tray support calculations. The letter requested that Florida Power Corporation (FPC) take appropriate steps to address the identified issues and inform the NRC Staff of the planned
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corrective actions. The purpose of this letter is to :;nbmit the FPC response to these identified issues. ;
1 In response to previous NRC Staff concerns, FPC made commitments in References 2 and 3 to perform Audits, Self-Assessments, and Third Party Reviews of the CR-3 USI A-46 program. )
These reviews are continuing.
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9901060089 981231 PDR ADOCK 05000302>
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CRYSTAL RIVER ENERGY COMPLEX: 15760 W. Power Line Street
- Crystal River, Florida 344284708 + (352)7954488 A FlorMs Progress Company
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' U.S. Nuclear Regulatory Commission )
Page 2 of 2 ,
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,, _ FPC pers$nnel discussed the issues that remain for resolving USI A-46 at CR-3 with the NRC Staff in a telephone conference call conducted on October 15,1998. This letter documents the FPC positions on the issues discussed during that conference call.- f FPC has previously committed to complete activities to resolve USI A-46 prior to restart from the Cycle 11 Refueling Outage, scheduled for Fall 1999. FPC considers it more appropriate to_.
identify a specific completion date for A-46 resolution. Accordingly, FPC will complete all actions specified in this submittal and in References 2 and 3 by December 15,1999.
j .
- If you.have any questions regarding this submittal, please contact Mr. Sid Powell, Manager,
- -Nuclear Licensing at (352) 56324883.
- c Sincerely, .
1 i
i Daniel L. Roderick, Director - I e Nuclear Engineering and Projects ;
+ DLR/rer/cgp cc: Regional Administrator, Region II ;
, Senior Resident Inspector NRR Project Manager
. Attachments A .' List of Regulatory Commitments B. Response to the NRC Audit Report Observations [
, C. Response to Appendix A of the NRC Audit Report D. Response to Cable Tray Calculation Concerns
- E.. ~ References P
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U.S. Nuclear Regulatory Commission - Attachment A !
Page1 of1-
. 3F129841-
-ATTACHMENT A LIST OF COMMITMENTS The following table identifies those actions committed to by Florida Power Corporation in this .
document. Any other actions discussed in the submittal represent intended or planned actions l
by Florida Power Corporation. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Manager, Nuclear Licensing of.any-questions regarding this document or any associated regulatory commitments.
Commitment Due Date FPC will . complete all - actions required to address the December 15,1999 outstanding issues for USI A-46 at Crystal River Unit 3 as specified in Attachments B, C, and D of this submittal.
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i U.S. Nuclear Regulatory Commission Attachment B !
3F1298-01 Page 1 of 9 ATTACHMENT B RESPONSE TO THE NRC AUDIT REPORT OBSERVATIONS Overview NRC to Florida Power Corporation (FPC) letter dated April 10,1998 (Reference 1),
forwarded the results of the NRC Staff audit of the Crystal River Unit 3 (CR-3) USI A-46 program. The letter requested that FPC take appropriate steps to address the issues identified in the enclosures, and inform the NRC Staff of the planned corrective actions. Enclosure 1, Attachment 2 of the letter, provided the NRC Audit Report. NRC Audit Report Section 3.0,
" Observations," identified audit findings regarding program elements that remain to be resolved. This attachment provides FPC response to these findings.
- 1. Overall USI A-46 Program Implementation (NRC Audit Report, Section 3.1) ,
. Issue:
The NRC Audit Report stated that the CR-3 Plant Specific Procedure (PSP) did not incorporate the NRC Staff's position as delineated in the NRC to FPC letter dated April 12, 1994 (Reference 2), and in the subsequent Safety Evaluation Report (SER) (Reference 3).
Resolution Strategy:
The NRC Audit Report stated that FPC had committed to reassess its position regarding completeness of the PSP, and to submit the results of its assessment and projected actions to address areas that may require further examination and evaluation prior to Refueling Outage 11. FPC will perform the additional actions to address deficiencies in the PSP, as identified by the NRC, as described in detail in the following sections. FPC will complete these actions by December 15, 1999. These issues and FPC planned actions were' discussed with the NRC Staff during the October 15,1998 telephone conference.
- 2. Status of USI A-46 Program (NRC Audit Report, Section 3.2)
Issue:
The NRC Audit . Report stated that FPC had committed to provide a schedule for final resolution of the remaining outliers from the FPC seismic walkdowns performed using the PSP before startup from Refueling Outage 11. The NRC Audit Report also stated that the observations and equipment-specific findings that were noted during the audit will need to be addressed by FPC for resolution. Appendix A of the NRC Audit Report further described the equipment-specific findings.
Resolution Strategy:
Previously. FPC had stated in letter to NRC dated December 16,1997 (Reference 4), that all remaining outliers identified during the oiiginal walkdowns based on the PSP will be resolved before restart from Refueling Outage 11, scheduled for Fall 1999. Those outliers, as well as any additional outliers which are identified by the additional evaluations, calculations, drawing i
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U.S. Nuclear Regulatory Commission Attaclunent B 3F1298-01 Page 2 of 9 and photograph reviews, and walkdowns to address the deficiencies in the PSP will be resolved by December 15,1999. ;
Appendix A of the NRC Audit Report identified fifteen equipment-specific findings regarding seismic qualification. Attachment C of this submittal addresses these findings.
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- 3. Qualification of Review Engineers (NRC Audit Report, Section 3.3)
Issue:
The NRC Audit Report stated that, in certain aspects, adequate interaction between the seismic review engineers and the systems engineers may have not taken place during the seismic adequacy verification process (examples are discussed in Section 3.9 of the NRC Audit Report).
Resolution Strategy:
The examples discussed in Section 3.9 of the NRC Audit Report involve the completeness of the CR-3 Safe Shutdown Equipment List (SSEL). These examples, and the FPC response to the equipment-specific concerns noted in the NRC Audit Report involving the SSEL, are further discussed in Section 9 below.
FPC has contracted Duke Engineering and Services to perform an independent assessment of the SSEL. This independent assessment is to ensure the SSEL is complete in representing the required safe shutdown equipment, and that the equipment on the SSEL is properly classified.
- 4. Responses to Request for Additional Information (RAI), (NRC Audit Report, Section 3.4)
Issue:
The NRC Audit Report stated that FPC did not address the open issues identified in the SER (Reference 3), and that the response to many of the RAI items (Reference 5) was still incomplete at the t!me of the NRC audit. The NRC Audit Report also stated that FPC submitted additional information concerning these issues (Reference 4), and that this additional information is under staff review and will be addressed in the final SER for closure of USI A-46 at CR-3.
Resolution Strategy:
The additional information provided by FPC (Reference 4), as supplemented by the information provided in this submittal, fully addresses the open issues identified in the SER and the remaining RAI items identified in the NRC Audit Report.
~ U.S. Nuclear Regulatory Commission Attachment B 3F1298-01 ~ Page 3 of 9 5.. Anchorage Verification (NRC Audit Report, Section 3.5)
Issue: l
- The'NRC Audit Report stated that an adequate anchorage verification process should consist of three major steps:
- 1) verify that equipment is anchored; j 2)' verify 'that field installation followed standard engineering practice; and- l j
3)- verify that the as-installed anchorage configuration is capable of withstr.4 ding and
- transmitting the seismic load. j The NRC ' Audit Report also stated that the NRC audit team identified some instances of
, anchor installation deficiencies and conditions involving unverified anchorage strength in !
1 L accordance with Generic Implementation Plan, Revision 2 (GIP-2) guidelines. It was also I ' stated that, for inadequate, improper or unconventional anchorage installations, it was
- questionable whether the FPC engineers could judge anchorage adequacy for the seismic
' demand without performing some calculations (e.g., following GIP-2 guidelines). The NRC :
' Audit Report reiterated the position delineated in the SER (Reference 3) that FPC should reassess its anchorage program following the guidance provided in Section 4.4 of GIP-2.
i Resolution Strategy:
The NRC Audit Report acknowledged the commitmeat to perform GIP-2 anchorage l calculations for approximately 50% of the electrical equipment on the SSEL before restart from Refueling Outage 11. FPC estimates that approximately 100 component anchorage calculations will be required to satisfy this commitment. FPC committed to perform these i
additional. calculations in Reference 4. FPC will complete these anchorage calculations by ;
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December 15,1999.
The choice of this sample size is not based on statistical principles. Rather, it is an estimate of i the population of equipment required to be included in these calculations to ensure all
- electrical equipment on the SSEL is adequately represemed. The selected equipment will ,
include equipment groups that can be qualified by typical calculations, and equipment for _l which calculations performed will bound similar equipment that would experience a lesser seismic demand. Equipment with anchorage that cannot be qualified as a part of the above '
equipment groupings will be considered unique, and will be evaluated separately. Therefore, the approximately 100 pieces of equipment will be selected to represent or bound all of the j electrical equipment on the SSEL.
GIP-2 methodology will be used for these calculations. The calculations will include review of the Screening Evaluation Work Sheets (SEWS) for all of the electrical equipment on the i
.SSEL -to identify the limiting cases of anchorage type, bolt pattern, and uniqueness. .
. Equipment will' be. selected. that bounds similar equipment of a lesser seismic demand.
LWalkdowns and inspections will be performed as required to complete and verify these
> calculations. ,
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U.S. Nuclear Regulatory Commission Attachment B 3F1298-01 Page 4 of 9.
- 6. _ Meeting All GIP-2 Caveats (NRC Audit Report,' Section 3.6)
' Issue: ,
The NRC Audit Report stated that meeting all GIP-2 caveats is a prerequisite for use of the i earthquake experience-based approach as pointed out in the SER (Reference 3). It is also i stated that the NRC Staff does not accept prescreening of caveats based on the argument of
. low seismicity. The_ NRC Audit Report stated that one of the weaknesses in the PSP is that it i does not require the engineers to adhere to the entire GIP-2 caveat list. ' The NRC Audit
' Report stated that the issue of missing caveats can be resolved when FPC confirms that all GIP-2 caveats were systematically verified and satisfied for all safe shutdown equipment.
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- Resolution Strategy:
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! The NRC Audit Report acknowledged that FPC had submitted additional information in Reference 4~ regarding. omission of certain GIP-2 caveats from the PSP, and that FPC had
- Jagreed to reassess this issue. In order to resolve this issue, FPC will perform additional' evaluations to systematically address GIP-2 caveats that had been previously prescreened.
! Any additional outliers identified by these evaluations will be resolved by December 15,1999.
The following list identifies GIP-2 caveats that were prescreened (not included in the PSP).
They are generically grouped by type of caveat rather than by equipment class.
Attached Weight of100 Pounds or Less i Cutouts Not Large Doors / Buckets Secured Natural Frequency Relative to 8 Hz Limit Considered Valve Body Not of Cast fron Valve Yoke Not of Cast fron Valve Operator Cantilever Length V (Fluid-Operated Valves) Mounted on One-Inch Diameter Pipe or Greater
~ - Actuator and Yoke Not independently Braced
'No Possibility of Excessive Duct Distortion Evaluate Computers and Programmable Controllers Separately Evaluate Strip Chart Recorders Separately c No Reliance on Weak-Way Bending of Steel Plate or Structural Steel Shapes Contains Only Circuit Breakers and Switches Check ofLong Unsupported Piping FPC will implement the following actions to address each of the above GIP-2 caveats, and will summarize the results in the appropriate SEWS packages, A. Caveats that were implicitly included in original walkdowns:
Doors / Buckets Secured Evaluate Computers and Programmable Controllers Separately Evaluate Strip Chart Recorders Separately
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U.S. Nuclear Regulatory Commission Attachment B'
- .3F12984)1 Page 5 of 9 ;
i Contains only Circuit Breakers and Switches .
t . The walkdowns performed in 1995, in accordance with Revision 1 of the CR-3 PSP, required '
' that all cabinets on the SSEL be opened and inspected by experienced and trained industry experts, referred to as Seismic Capability Engineers (SCEs). Although these caveats were not specifically noted. on the SEWS, these experts were' knowledgeable of these caveats and considered them during the walkdowns. ,
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~ FPC will obtain a signed letter from each SCE documenting that these caveats were considered
' as part of the walkdowns. These letters will be placed in the SEWS files. The signed letters l will be' formal acknowledgment by the SCEs that these caveats were implicitly reviewed.' In addition, FPC will' perform a walkdown of a sample of these caveats to ensure there are no ,
' concerns. The sample will consist of at least ten percent of the associated population. If any .
- outliers are identified in the population that should have been identified during the original '
L walkdowns, then the entire population will be re-evaluated. :
' Based on the above, FPC concludes that the walkdowns and reviews performed previously l
- adequately address these caveats. s
- B; Not applicable to specific equipment at CR-3
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c Natural Frequency Relative to 8-Hz Limit Considered This GIP-2 caveat was reviewed during preparation'of the CR-3 Technical Basis Document. ;
A comparison of CR-3-plant specific response spectra was made against GIP-2 reference !
spectra All electrical cabinets on the SSEL are located at floor elevations where the specific !
CR-3 response spectra are below GIP-2 reference spectra (regardless of frequency). As a result, the 8-Hz issue.does not apply for these cabinets. Based on this, FPC concludes that this caveat has been adequately addressed. ,
C. Additional Walkdowns:
L
. Attached Weight of100 Pounds or Less l'
' Cutouts Not Large^
No Possibility of Excessive Duct Distortion
- Check ofLong Unsupported Piping ,
FPC will perform an additional walkdown to address these GIP-2 caveats. This walkdown !
-will address the specific caveats for the affected equipment classes that were previously excluded from consideration. Appropriate notes, observations, and assessments will be made
. b document the results of these walkdowns. >
D. Engineering Evaluations and Walkdowns:
Valve Body Not of Ost fron Valve Yoke Not of Cast fron
[ . Valve Operator CantileverLength U Mounted on One-inch Diameter Pipe or Greater i
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, J U.S. Nucle'ar Regul: tory Commission- Attachment B L '
'3F1298 Page 6 of 9 .
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,' ' . Actuator and Yoke Not independently Braced -
FPC .will perform a search of the CR-3_ Configuration Management Information System,. '
review drawings and photographs, and perform additional walkdowns as necessary to address
- - these GIP-2 caveats. .
- E. Engineering Evaluations / Calculations:
No Reliaisce on Weak-Way Bending of Steel Plate or Structural Steel Shapes
-
- FPC will examine photographs and perform a bounding calculation to. address this GIP-2 i
- caveat. 1
- 7. ' Operator Action / Emergency Lighting (NRC Audit Report, Section 3.7)
- 3. .
. -Issue: ,
[ (The NRC Audit Report stated that the NRC Staff was in general agreement with FPC that falling or failure of components during a seismic event is not a major concern for maintaining .
a clear path through the Turbine Building. However, the NRC Audit Report also stated that F the NRC audit team did not find any reliable emergency lighting system that was seismically .
[ . verified, and that the only light available would be flashlights or similar portable lights. -
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' Resolution Strategy: :
i FPC will evaluate all of the plante emergency battery-powered lighting to verify it is seismically rugged. Any deficiencies will be evaluated and resolved to ensure operator actions i in . Abnormal Procedures (APs) or Emergency Operating Procedures (EOPs) are not impacted.
.The impact review will be conducted in accordance with the AP/EOP verification and
_ validation process. ' One of the requirements of this verification and validation process is to ,
i ensure that adequate lighting is available for performance of required operator actions. This I lighting evaluation considers- both normal and loss-of-offsite power (LOOP) conditions, e including possible reliance'on emergency battery-powered lighting, or the use of hand-held ;
l . lighting.
, Hand-held lighting is staged in dedicated EOP boxes that are strategically located throughout the plant. To ensure the n.cessary tools and materials are available to perform required AP p ' and EOP. operator actions, a surveillance is performed to. verify the contents of each EOP box e on'a periodic basis. Plant operators have received training on the use of the staged hand-held
' lighting; 'fhis training. includes use of the different types of hand-held lighting during a simulated IOOP condition, and use of hand-held lighting in conjunction with the use of ladders and manipulation of valves. The AP/EOP verification and. validation process and the emergency lighting' walkdowns provide reasonable assurance that sufficient lighting will be
.available for safe plant shutdown following a seismic event. Any actions taken after a seismic event that are not specified in the APs or EOPs would not be required for plant safety, and do
- not require the same level of verification and validation.
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- ((. / W L U.S. Nuclea'r Regulatory Commission
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Attachment B ;
3F1298-Ol' Page 7 of 9 ?
- 8. ' Relay Review (NRC Audit Report, Section 3.8) ;
' Issue:
- The NRC AuditL Rep $rt stated that the NRC' Staff.had accepted-in its SER (Reference 3) a -- i
_ a. reduced scope review of relays, which includes ~ replacement or otherwise justification for the
' Bad Actor" relays and spot checks of mounting of other relays. The NRC Audit Report also: .!
stated that FPC determined that none of the'" Bad Actor" relays would be required to perform any safety function. However, the NRC Audit Report stated the SER allowed a reduced scope :
review with the unde'rstanding that FPC would develop a " top-level" procedure to deal with.
. potential relay chatter. ,
Resolution Strategy: -
7' FPC has revised AP-%1, " Earthquake," to address the potential for relay chatter and to
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L discuss operator actions in response to relay chatter. As stated in the NRC Audit Report, FPC i
- had submitted AP-%1,- Revision 8, as a " top-level" procedur~ e that was under review by the-
' NRC Staff (Reference 6). ,
t The revision to AP-%1- was communicated to thefoperating crews through the Operator )'
Training Program as required reading. The intent of the revision was communicated through an Operations Study Book entry, which is a form of required reading. AP-%1 will also be ,
reviewed periodically by the operators in accordance with the Operator Training Program l guidelines. !
I Based on this, no further ac'tions are required at this time to address relay review.
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- 9. Identification of Safe Shutdown Equipment (NRC Audit Report, Section 3.9) - l Issue:
i The NRC Audit Report stated that the NRC audit team had a concern that the SSEL was
, determined - without adequate interaction. among engineers with the needed collective ,
background. ' The NRC Audit Report also stated that this lack of strong engineering interaction :
was further. supported by observation during the field inspection when the NRC audit team i
-- found examples of mappropriate equipment class identification.
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- Resolution Strategy: l t
1 The NRC Audit Report stated that FPC agreed to take the following actions: i
> 1. . FPC will re-examine .. the completeness ' of. the SSEL- by involving engineers with i backgrounds in all aspects of safe shutdown path identification and equipment functionality 1
,/ . to confirm that no' equipment items need to be added to the current SSEL list. !
- 2. FPC will verify whether all equipment items are appropriately classified. Reevaluation t will be necessary for mappropriately classified equipraent.
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' f 3. . FPCJwill verify whether.the diesel grounding resistor needs to be added to the safe
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shutdown list. If so,' the adequacy. of the ceramic insulators in the load path should be - l
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i U.S. Nuclear Regulatory Commission Attachment B I 3F1298-01 Page 8 of 9 investigated. The NRC Audit Report also stated that FPC submitted related information by f letter dated December 16,1997 (Reference 4), and will be reviewed by the NRC Staff. l To address Item 1, FPC 'has contracted Duke Engineering and Services to perform an ,
independent assessment of the SSEL. This independent assessment is to ensure the SSEL is complete in representing the required safe shutdown equipment, and that the equipment on the SSEL is properly classified. The independent assessment will verify that the methodology l' used to review the safe shutdown paths and equipment at CR-3 is consistent with Section 3 of GIP-2, and will be completed by December 15, 1999. j Attachment H of the FPC submittal dated December 16,1997 (Reference 4) previously j addressed Item 2. The classification of equipment on the SSEL will be verified to be l consistent with the equipment classifications identified in Table 3-1 of GIP-2 during the Duke Engineering and Services independent assessment.
t Attachment G of the FPC submittal dated December 16,1997 (Reference 4) previously r addressed Item 3. A further discussion of this issue is inchJed in Attachment C of this l submittal under NRC Issue Number 12, and concludes that the diesel grounding resistor does - :
not need to be placed on the CR-3 SSEL. ,
- 10. Equipment Internals (NRC Audit Report, Section 3.10) !
Issue:
The NRC Audit Report stated that the verification of seismic adequacy with respect to the ,
equipment caveats requires that FPC seismic engineers open electrical cabinets to verify adequate mounting of internals and confirm that there are "no other concerns." The NRC i
Audit Report stated that the NRC audit team had identified two specific examples where the seismic adequacy of internals was questionable.
Resolution Strategy:
Attachment I of the FPC submittal dated December 16,1997 (Reference 4) previously addressed these two specific examples, and provided technical justification for the !
determination'that both of these specific equipment issues are acceptable without any further l action. A further discussion of these issues is included in Attachment C of this submittal -
i under NRC Issue Numbers 1 and 3.
l 11. Supports for Cable Trays and Conduits (NRC Audit Report, Section 3.11)
Issue:
l ;
The NRC Audit Report stated that the supports for cable trays and conduit were not evaluated ;
by use of GIP-2 or similar experience-based approach. The NRC Audit Report also noted ;
) FPC statements during the NRC atsdit that these support systems were analyzed and designed ;
for lateral as well as vertical seismic loads in accordance with the design basis for CR-3. !
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' U.S. Nuclear Regulatory Commis'sion Attachment B j 3F1298-01 Page 9 of 9 - ,
Resolution Strategy:
FPC will verify that' bounding documentation is available to show that all raceways required , l
-for safe shutdown of the plant are seismically adequate. Specifically, existing design basis i calculations and walkdown information will be reviewed to ensure that it provides assurance ,
equivalent to that provided by GIP-2 methodology. -In cases where the existing documentation ,
. does not. provide thi.c level.of assurance, additional walkdowns and/or calculations will be
, performed. .
- 12. Tanks anil Heat Exchangers (NRC Audit Report, Section 3.12) ;
Issue:
The NRC Audit Report stated that the NRC audit team had identified cracks at the base of the - f
- Emergency Feedwater Tank (EFT-1) support pad that were not previously identified by FPC. !
The NRC Audit- Report also' stated that FPC had declared the Condensate Storage Tank i
'(CDT-1) to be an outlier, and noted that a calculation is required to close the issue. {
Resolution Strategy: j The NRC Audit Report stated that FPC had evaluated the cracks in the EFT-1 support pads, and FPC had determined that they are confined to the grout only, and do not affect the i integrity of the stmetural concrete foundation. - As noted in the NRC Audit Report, FPC has submitted the seismic evaluation of the tank itself (EFT-2) to the NRC (Attachment B of Reference 4), which is currently under NRC Staff review. ;
FPC is completing a calculation for CDT-1 that will demonstrate that the tank is seismically l qualified. This calculation will be submitted to the NRC for review by December 15,1999. :
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o U.S. Nucl:ar Regulatory Commission Attachment C 3F1298-01 Page 1 of 13 1
ATTACHMENT C RESPONSE TO APPENDIX A OF THE NRC AUDIT REPORT Appendix A of Reference 1 discussed fifteen concerns regarding specific equipment items in the CR-3 USI A-46 program. This attachment discusses how FPC has addressed each of these concerns. The following table identifies the equipment items discussed in Appendix A of Reference 1 and this attachment.
NRC Issue FPC FPC Number Tag Number (s) Tag Number Description 1 MTSW-3A, 3C 480V Turbine Auxiliary Bus 3A and 480V Reactor Auxiliary Bus 3A 2 DPDP-5B Engineered Safeguards DC Panel 3 ATCP-1 Anticipated Transient Without Scram (ATWS)
Logic Cabinet 4 MTXS-1 Manual Transfer Switch For MCC 3AB, MTMC-7 5 MTSW-3F Engineered Safeguard Bus 3B South Section 6 VBXS-1 A,1C Static Switch With Auto-Retransfer 7 VBXS-1E Manual Transfer Switch 8 AHCP-4 Silicone Controlled Rectifier Cabinet For AHHE-4A & 4B 9 AHP-1 A, IB,1C, ID Automatic Temperature Control Air Compressor 10 CCBT Central Control Board Termination Cabinet 11 EGDG-1B Emergency Diesel Generator 1B 12 (N. A.) Diesel Grounding Resistor 13 MTMC-18 480V Reactor MCC 3A2 14 DPXS-1 Manual Transfer Switch For Power Select To MUP-3B & MUP SB 15 RWP-1, 2A, 3A Normal Nuclear Services Seawater Pump, Emergency Nuclear Service Seawater Pump, i DH Service Seawater Pump NRC Issue Number 1:
Low Voltage Switchgear MTSW-3A, 3C The transformer bay of the lineup (3A) was opened. Several ceramic insulators were observed. Since ceramic is brittle, the need to assess their adequacy was noted to the licensee.
No cabinet anchorage was observed. In another location (3C), plug welds were observed as the means of anchorage. Their strength tofully transmit the seismic loads within allowable stress limits nus considered questionable. Sources of potential installation deficiencies for plug welds are asfollows:
e inadequate fusion of the base material of the underlying component (e.g., plate, angle, channel, etc.).
o-Lack offusion in the base metal around the hole.
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- U.S.- Nuclear Regulatory Commission - Attachment C '
3F1298-01' Page 2 of 13 e Burning of connecting component around the hole, especially if the metal is thin.
The above' concerns are genericfor electrical equipment supported by means ofplug welds. i FPC Response:
The' concern noted above for the ceramic insulators was addressed previously in Attachment I P of Reference 4. -The following.information is excerpted from Attachment I with additional ,
, information provided where appropriate.
MTSW-1A is the 480V Turbine Auxiliary Bus and MTSW-1C is the 480V Reactor Auxiliary Bus. These components are located in the Turbine Building, Elevation 95'. FPC identified i
- these components as outliers during the 1995 walkdowns because of inadequate anchorage. To f address this, FPC will add additional cabinet base weldments and bolting. This modification ,
. will be completed by December 15,1999. :
LFPC corisulted with industry experts to further address the issue on the ceramic.insulato:s at
- the bus bar. Based on the input from these industry experts, FPC concluded that the ceramic 1 insulators would be acceptable, provided the insulators are not particularly long and slender ,
and that the load path does not induce any great beixting moments into the insulators. The ;
experts acknowledged there is no exact acceptance criteria, only that the SCEs shall use their collective judgment to determine the .:ceptability of the ceramic insulators, based on their i
length, diameter, and load path.
FPC SCEs have inspected the interior of the cabinets. The ceramic insulators are used to ,
' insulate the bus bars. The length (<4") is not excessive.' The diameter (> l 1/2") is not ,
particularly small. The primary loading on these insulators would be compression. The load path does not appear to substantially load these insulators in any type of bending moment.
Therefore, based on the judgment of the SCEs, the short and robust configuration of these insulators is acceptable. )
GIP-2 does not indicate ceramic insulators are an issue with respect to Class 2 equipment. !
Two separate SCE inspections (original in accordance with the' CR-3 PSP and follow-up in
~ accordance with GIP-2) did not identify these specific insulators as a seienic concern.
~
FPC analysis / calculation S97-0541 (Reference 8) envelops the electrical equipment anchorage l issue. This calculation was included in Attachment A to Reference 4. Attachment E of
! calculation S97-0541 includes a calculation. for " typical" low voltage switchgear. This- l calculation included calculations for spot welds. The qualification of the spot welds was based ;
on GIP-2 and EPRI NP-5228. The calculation showed the qualification of spot welds to be l
' acceptable. l Based on meeting GIP-2 acceptance criteria and the evaluations discussed above, FPC concludes the anchorage and ceramic insulators are acceptable.
l
- J
1 a U.S. Nuclear Regulatory Commission . Attachment C 3F1298-01 Page 3 of 13 l
[
NRC Issue Number 2:-
Distribution Panel DPDP-SB i- Only the rear side of this two-bay floor-mounted cabinet waJ externally, visually inspected.
l There was a large gap (approximately 3/4 inch) between the two structural components
. connected by the base anchor bolts (appears to be 3/8 inch diameter). This was considered to I
be an example of a questionable and unconventional anchorage installation. The front side
! could not be inspected, so the overall effect on the equipment anchorage strength could not be-judged The licensee's screening evaluation work sheets (SEWS) do not seem to indicate any
- l. anchorage deficiency.
I j FPC Response:
J This is a 250/125V DC Engineered Safeguards (ES) Distribution Panel lo::ated in the Control Complex, Elevation 124' (See Figure 1 below). FPC Engineering re-inspected DPDP-5 and l
[
did not identify any large gaps between components, floor, or wall. Thus, it appears the NRC i Audit Report concern does not apply to DPDP-5B. FPC concludes that DPDP-5B meets all
- required acceptance criteria.
FPC SCEs inspected equipment in the same room with DPDP-5B and in adjacent areas to
- identify any equipment that displayed gaps as described above. FPC was unable to identify j any such equipment.
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.U.S. Nuclear Regulatory Commission Attachment C 3F1298-01 Page 4 of 13 NRC Issue Number 3:
I&C Panel, ATWS Logic Cabinet ATCP-1
~A box-type device about 6-inches high and-1-foot long was supported as a cantilever byfour approximately 1/8 inch diameter screws (slotted holes on one side). The staff audit team Judged that further investigation was necessary to confirm the structural adequacy of the connection.
FPC Response:
_.This component is the Anticipated Transie'nt Without Scram (ATWS) Logic Cabinet. This cabinet is located in the Control Complex, Elevation 124', Emergency Feedwater Initiation and Control (EFIC) Room C. FPC interprets this concern to be that one of the logic boards had an excessive cantilever length off the main vertical support frame'.
Attachment I of Reference 4 discussed that field measurements show the logic board to be 11"
- deep x 171/2" wide x 7" high. The board is attached to the metal uprights by four 1/8"
- diameter screws. The screws are spaced 4" on-centers. The screws are mounted into drilled and tapped holes on the uprights. The upright members themselves appear robust and are judged to be adequate for the bending loads.
An engineering evaluation of the subject logic board _was performed. The evaluation showed
'that the screw capacity, under seismic loading, is adequate to support the cantilever length.
The detailed evaluation was submitted to the NRC as part of Attachment I of Reference 4.
Based on the engineering evaluation, FPC concludes this equipment is adequate.
NRC Issue Number 4:
Low Voltage Switchgear MTXS-1 The door of the cabinet was found to be extremely flexible and might rattle during an earthquake. Also, the rear side of the tall (over 6 feet) cabinet was too close to the wall (less than l '4 inch). Thus, spatialinteraction may not be acceptable. Furthermore, it was unclear .
whether this equipment piece belongs to the low voltage switchgear class as identified in the SSEL.
FPC Response:
This component is the "ES MCC 3AB Input Power Transfer Switch." It is located in the
. Control Complex, Elevation 124'. FPC engineers have reviewed the applicable vendor cabinet drawing and have also performed a supplememal walkdown. The following comments are based on this review.
The cabinet is a NEMA 1 cabinet fabricated from 10-gauge sheet steel. This sheet steel is welded to internal steel frame members that consist of 3" x 5" x 1/4" vertical angle stiffeners in the back corners and 1 1/2" x 1 1/2" x 1/4" vertical angle stiffeners in the front corners.
The base and top members also consist of angle steel. This is a rugged cabinet. A tug test
- was performed in.the field and.no displacement was observed. In the judgment of the
i J.S. Nuclear Regulatory Commission Attachment C
, 3F1298-01 Page 5 of 13 l
walkdown team members, the tug test simulated the force that might be experienced during a seismic event. Based on the,walkdown, the drawing review, and the tug test, FPC considers .i
- the cabinet frame as robust. FPC considers the displacements that might occur during a l
- seismic event to be small.
] The gap between the cabinet and wall was noted previously on the SEWS. The comment from
! 'the SEWS, dated 3/121%, is: ' Top of cabinet 1/2"from wall - bottom 3ll6"from wall.
- Judged not to be a credible interaction. ' The SEWS also notes that a tug test was performed i on the cabinet and that the anchorage was concluded to be seismically adequate. .This was considered to validate that the gap is acceptable. ;
l ,
- _ FPC Stru tural Engineering personnel examined the door. There is approximately 1/8" to 1/4" of play in the two half doors where they meet at the center. The door halves are fastened. GIP-2 caveat would verify the doors are fastened to prevent opening and possibly ;
l ::reating tepeated impacts against the housing. The cabinet is well anchored as discussed ;
j above and the seismic demand at the 124' Control Complex level is low. Therefore, while the
[
potential does exist for the doors tp have minor movement while still fastened, the chance that
- this might cause relay chatter is remote.
In addition, in the remote chance the internal relays did chatter, FPC operators will recognize I and take. appropriate actions to correct any conditions resulting from the relay chatter. These j actions are discussed in CR-3 Abnormal Procedure AP-961, " Earthquake." AP-%1, 4
Revision 8, was submitted to the NRC by Reference 6. The NRC has agreed to this approach
- in their SER, dated April 12,1994 (Reference 2). Therefore, FPC concludes the door to be acceptable.
The concern regarding the equipment classification has been previously addressed in '
i Attachment H of FPC letter to NRC dated December 16,1997 (Reference 4). MTXS-1 was l discussed in that attachment and was found to be correctly classified. FPC has contracted i Duke Engineering and Services to perform an independent assessment of the SSEL. This l
independent assessment is to ensure that the equipment on the SSEL is properly classified.
{. NRC Issue Number 5: '
Low Voltage Switchgear MTSW-2F ;
It was unclear whether the added weight of the transformer mounted on top of the cabinet i exceeds the limit recommended in GIP-2. The caveat on limitation of external weight was not .
included in ' the PSP. Further, the reliability of plug welds as anchorage needs to be
- investigated (see Item No.1 above, for discussion ofplug welds).
- FPC Response
I This component is the 4160V ES Bus 3B switchgear cabinet. It is k>cated in the Control
[ Complex, Elevation 108'. This concern has previously been addressed in Attachment I of
- Reference 4.
The NRC Staff clarified this comment during the November 1997 audit. The NRC concern is
- - that the' potential transformer unit is in a separate cabinet bolted to the top of the main 1
U.S. Nuclear Regulatory Commission Attachment C 3F1298-01 Page 6 of 13 ,
i switchgear cabinet. This arrangement is typical of other ES switchgear cabinets located in the Control Complex. ,
The PSP SEWS mentions the potential transformer cabinet, but not as an outlier for this concern. The transformer cabinet was opened and examined as part of the walkdown. The SEWS states that the cabinet in question is rigidly attached to the main host cabinet and all internal components are adequately attached. 1 The SCE who did this walkdown acknowledged the cabinet is external to the main cabinet.
Ilowever, no specific discussion is made on the weight of the transformer cabinet. !
Regardless, the SCE judged the component to be acceptable. j GIP-2, Appendix B, states the following for this caveat: !
"The concern is directed primarily for equipment not normally supplied with the switchgear i and thereby possibly not included in the earthquake experience equipment class. The load ;
path of the attached component through the cabinet should be carefully examined. In addition, i
~ its attachment should be reviewed to ascertain whether the attached component may become a seismic interaction hazard source. "
The 4160V ES Switchgear cabinets are safety-related, Seismic Class I components, and the :
existing plant drawings show that the potential transformer is part of original design. The external cabinet is anchored well to the main cabinet, and the internals of the host cabinet are i judged to provide an adequate load path to the anchorage. 3 The weight of the trans~ormer cabinet cannot readily be determined. However, based on the j above discussion the switchgear assembly is considered acceptable and meets the intent of the i GIP-2 caveat. ;
The issue of spc4 welds was addressed earlier under NRC Issue Number 1 above. FPC analysis calculation S97-0541 was intended to be a bounding anchorage calculation for electrical equipment. This calculation determined allowable loads for spot welds and qualified J the anchorage of a " typical" switchgear. As committed to in Reference 4, FPC has agreed to i perform additional anchorage calculations. These calculations will more fully verify the adequacy of any anchorage using spot welds.
NRC Issue Number 6:
- 'L.ow Voltage Switchgear VBXS-1A, JC L This is a wall-mounted box-type panel 16-inches de
- p that looked like a distribution
[ !
panelboard although the SSEL identiped the cabinet as a low voltage switchgear'(Class 2, j i Table 5-6 of FPC Summary Report). It was noted that the depth of a typicalpanelboard in the !
' database is 6 to 12 inches according to GIP-2, and the 16-inch depth may make this j panelboard an outlier. The licensee was informed to consider the reevaluation of this item of \
- equipment by use of the applicable caveatsfor a distribution panel, i
e
._ ~ .; _ _ _ . _ = - .
i U..S. Nuclear Regulatory Commission Attachment C
- Page 7 of 13 13F1298-01 '
FPC Response: !
The question; about the equipment classification ' has ..been previously addressed in ;
Attachment H of Reference 4. This document reflects the FPC conclusion tha* the equipment l lis correctly classified as' a, Class 2 component. FPC has contracted Duke Fagineering and Services to perform an independent assessment of the SSEL, which will provide ad!itional :
assurance that the equipment on the SSEL is properly classified. GIP-2 does not identify a i caveat related to component depth'for Class 2 components.' Therefore, the 16-inch depth of !
the equipment is not a problem. Additionally, the electrical equipment anchorage calculations ;
to be performed,' discussed in the response to Item number 5 in Attachment B, will bound the 'j anchorage of this component.-
NRC Issue Number 7: i
^ Low Voltage Switchgear VBXS-1E
^
l Similar to Item No. 6, this wall-mounted panelboard was classified as low voltage switchgear. ;
. This was considered to be an inappropriate classfication. The depth of this panelboard is within the database and, therefore, is not a problemfor thispanelboard.
FPC Response:
The question about the equipment classification has ,been - previously addressed in l Attachment H of Reference 4. This document reflects the FPC conclusion that the equipment !
is' correctly classified. FPC has contracted Duke Engineering and Services to perform an !
independent assessment of' the SSEL, which will provide additional assurance that the !
equipment on the SSEL is properly classified.
- NRC Issue Number 8: ;
16 C Panel AHCP-4 i i
This three-bayfloor-mountedpanel could not be opened during the audit. An external visual j
. inspection revealed large cutouts on both the top and bottom of each bay. The loadpath may l need to be investigated. Theframe anchorage connection may also need to be investigated. A FPC Response:'
t This component has been reexamined. An anchorage deficiency was determined to exist at the j anchor bolt locations. This deficiency was determined to be a tear-out around the anchor bolt -l itself, since there was insufficient material strength around the thin sheet metal base to resist the local tearing force around the bolt. ' The deficiency was documented in accordance with the
> . FPC corrective action program. This cabinet was then declared an outlier, and a modification ,
package was developed to resolve the issue, t
' The concern regarding the large cutouts involves only the cutouts in the doors to the cabinet :
. (See Figure 2 below). Field inspection and review'of the "ROBICON" Vendor Drawing. ;
_ 1415-026 shows the cutouts to be 16" x 24" on the three doors to the cabinet. The doors are !
not part of the load path of the cabinet, and the cabinet has internal stiffeners. In addition, the
~
7 majorJ1oad components 'inside 'this ' cabinet Pre mounted on a separate,- internal frame that l i
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j U.S. Nuclear Regulatory Commission Attachment C .
s 3F1298-01 Page 8 of 13 '
(
} ,
j attaches directly to the base of the cabinet. Therefore, the cutouts mentioned above do not )
j affect the load path and are considered acceptable. l
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- NRC Issue Number 9
Air Compressors AHP-1A,1B,1C, and 1D l
l Thefront and rear ends of each compressor are anchored to separate pedestals through soft l pads. The potential of reladve displacement and its effect on equipmentfunctionality needs to be investigated.
l I FPC Response:
i These components are the Control Complex HVAC Air Compressors, located in the Control Complex, Elevation 164' (See Figure 3 below). They are all similar, and the discussion below
! is typical for all four components.
i I The above equipment is shown on " Johnson Service Company" Vendor Drawing
! 143-56-GP-3. The vibration isolation pads are shown on the drawing as a "Corfund" block j with dimensions of 4" x 4" x 1" thick. These pads are located under the support feet of the I compressor reservoir tank. The vendor drawing shows the support feet to be anchored with Rawl Expansion Anchors. The compressors are installed on raised concrete piers that are approximately 10" high, 8" wide, and 24" long.
4
! During walkdowns conducted in March,1995, the SCEs noted these components as outliers l because the anchorage and reinforcement of the concrete piers could not readily be identified.
- In September,1997, an ultrasonic test (UT) was performed on all the anchor bolts. The UT revealed that the compressors were anchored with threaried rods and not with the Rawl
! l I
f
i l U.S. Nuclear Regulatory Commission Attaciunent C i
3F1298-01 Page 9 of 13
) expansion anchors. The UT found that the threaded rods had lengths ranging from 15 3/4" to 171/4". These threaded rods are of sufficient length to not only anchor the equipment but to l
l extend directly into the concrete floor slab. These rods prevent lateral displacement of the j equipment.
The elastomeric isolators were reviewed with respect to Section 4.4 of GIP-2. Since the threaded rods prevent lateral displacements for all load cases (as described in GIP-2), then the vibration isolators are considered acceptable.
The four compressors consist of two air receiver tanks with two compressor units (motor and pump) mounted on top of each tank. Each air receiver tank is anchored into two concrete support piers by four support feet. The tank and the support feet are considered seismically rugged. The concrete support piers are relatively short, and are considered to be rigid. Little j displacement amplification will be experienced through these piers. Therefore, little !
j displacement is expected through the entire compressor assemblies. e i
l Since the anchor bolts tie directly into the floor slab system and the support piers are short, l there is very little potential for relative displacements at the support feet.
i
! Compressor AIIP-1D was included in the Stevenson & Associates walkdown. This walkdown f was performed in accordance with GIP-2 requirements and a SEWS sheet was completed j (dated 12/97). The SCEs noted the " neoprene" pads. Their evaluation shows the mounting to j be acceptable, i
I Considering the above information, FPC concludes the mounting of AHP-1A/1B/1C/1D is acceptable.
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, U.S. Nuclear Regulatory Commission Attachment C 3F1298-01 Page 10 of 13 NRC Issue Number 10:
I&C Panel CCBT This panel is adjacent but not connected to another panel whose anchorage could not be confirmed. The potential impact between the two panels and its effect on equipment functionality should be investigated.
FPC Response:
The condition described above was documented in the original SEWS for this equipment, ,
prepared in September 1995. The SEWS noted that the adjacent equipment (tagged TGAR) was anchored. Based on their experience and judgment, the walkdown team did not consider -
this condition to be a seismic concern.
'During the 1997 outage, a second walkdown effort was performed in accordance with GIP-2 .
requirements for selected equipment. This walkdown and assessment was conducted by
. Stevenson & Associates and is documented in Attachment A of Reference 4. This recent inspection noted this same condition. The following comment was made:
"CCBTframe structure very robustly anchored. Adjacent and in contact with cabinet TGAR; however, since CCBT contains no relays it isjudged OK. "
Tag number TGAR is not on the CR-3 SSEL. The cabinet labeled TGAR is the Turbine Generator Auxiliary Cabinet, and the FPC Configuration Management Information System (CMIS) identifies TGAR as non-safety related. ;
FPC seismic engineers examined I&C Panel CCBT on August 14, 1998 and confirmed that 7 there are no relays in it. In addition, there are no other seismically sensitive items in this cabinet.- The cabinet consists of terminal stiips and wiring, which is considered seismically insensitive.
Based on the above information, FPC concludes the initial walkdown findings are valid, and cabinet CCBT is acceptable. j NRC Issue Number 11:
Diesel Generator EGDG-1B The relay cabinet next to the diesel generator appears to be floor , or skid-mounted. It is bolted to a verticalplate, which in turn is welded to the skid. This load transfer system seems
. to be unconventional and may requirefurther investigation including its effect on relays within
' the cabinet.
FPC Response:
FPC Engineers have reviewed the mounting. The relay cabinet is bolted to three vertical bars approximately 1/2" thick x 21/2" wide. There are two bars at each end of the cabin t and one in the middle _of the cabinet. This makes for a robust composite frame. These three bars are ,
welded to.the bottom of the skid and again approximately at the top flange of the skid. The top weld location is approximately at the mid-point of the cabinet. Since the center-of-gravity
U.S. Nuclear Regulatory Commission Attachment C 3F1298-01 Page 11 of 13 (CG) of the cabinet is located at the approximate location of the top weld, there would be little bending in the bars. The welds join the cabinet to the very rigid diesel skid members. This implies very little amplification, or differential movement, between the two weld locations.
The welds are judged to be adequate. A tug-test was performed on the assembly. The force used during the tug test was not quantified but would simulate the forces expected during a seismic event. No discernible deformation was observed.
FPC concludes that the arrangement is seismically rugged and acceptable.
NRC Issue Number 12:
Diesel Generator Grounding Resistor There are ; ?" ceramic supports in the load path of the grounding resistor. Being brittle, ceramics arr .,pically a poor choice of material to transfer the seismic load. If the grounding resistor is requiredfor safe shutdown, it should be added to the SSEL and the reliability of the ceramic insulators should be investigatedfrom a seismic vieupoint. (Information concerning the grounding resistor was submitted with the licensee's December 16,1997 letter. It is under staffreview.)
FPC Response:
As stated above, information to address this particular concern was included in Attachment G of Reference 4. That attachment contains an engineering evaluation prepared by FPC Electrical Engineers. The evaluation concludes the grounding resistor is not required for operability of tl- diesel and therefore does not need to be placed on the CR-3 SSEL NRC Issue Number 13:
Motor Control Center MTMC-18 The front welds to the embedded steel appeared questionable on one side (right-hand side whenfacing the MCC) of the cabinet's lineup. The rear side could not be inspected. If these welds are in the load transfer path, the construction did not seem to meet acceptable industry standards, and the adequacy of the welds should be investigated. Furthermore, the MCC seems to be too close to the wall (a gap ofless than 1/4 inch at locations of metalprotruding
[from} the MCCframe for a spatial interaction concern). The credit for any restraints from top conduits needs to be carefully considered since such conduits are typically connected to thin sheet metal on top and these connections can break under seismic loads.
FPC Response:
Motor Control Center MTMC-18 is shown in Figure 4 below. MTMC-18 was one of the items included in FPC Calculation S97-0541, which was included in Attachment A of Reference 4. This calculation qualified the cabinet anchorage to GIP-2 guidelines. The as-built welding was used in the calculation. In addition, a weld strength reduction was used to account for the potential of a poor or missing weld. The calculation showed a factor-of-safety margin of 2.9 (calculated weld stress versus GIP-2 allowable stress). No credit was made in the calculation for the conduit.
U.S. Nuclear Regulatory Ccmmission Attachment C I
3F1298-01 Page 12 of 13 i
l The 1/4" gap was examined by the SCEs and judged not to be a credible hazard. FPC Engineering has reviewed the calculation and statement made by the original SCEs. The cabinet is well anchored at the base. Also, the conduit noted above will act to some degree to >
restrain the top of the cabinet from seismic motion. The conduits are large, numerous, and well supported (See Figure 4 below). Therefore, taking some credit for the conduits is j considered valid and acceptable. FPC recognizes that the thin sheet metal should be examined i if full credit is to be taken for the conduit. Ilowever, in this case, the conduit is not included j in the anchorage calculation. Rather, the conduit is considered only in limiting the deflection
! of the top of the cabinet. FPC concludes the thin sheet steel is not a concern for t!ns i consideration.
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NRC Issue Number 14:
i Low Voltage Switchgear DPXS-1 1 This is a wall-mounted panel that may belong to thc class of distribution panel; however, the
} equipment was identified in the SSEL as Class 2 (i.e., low voltage switchgear) and evaluated 1 accordingly. Thisfinding was noted to the licenseefor investigation.
1
! FPC Response:
- The question about the equipment classification has been previously addressed in
! Attachment H of Reference 4. FPC concluded the equipment is correctly classified as
- documented in that attachment. FPC has contracted Duke Engineering and Services to j perform an independent assessment of the SSEL, which will provide additional assurance that the equipment on the SSEL is properly classified.
i
U.S.' Nuclear Regulatory Commission : Attachment C -
3F1298-01 Page 13 of 13. i i
NRC issue Number 15:'. j
~
Vertical Pumps RWP-1, 2A, 3A j
. Anchorage between the concrete pad and the concretefloor could not be conprmed to ensure adequate load transfer (see RAI Item No.10, Reference 7). This finding was noted to the licenseefor investigation.
FPC Response:
The item referenced above is NRC's RAI, dated January 28,1997 (Reference 5). ' That RAI requested the following The Report on Page 36, Section'5.1.3 states that "all reinforced concrete pads are integrally \
attached to the concretefloors by dowel. " Please explain how this was venfied. .l i
1 FPC's response to the RAI was submitted by Reference 9. The answer was that the SCEs used plant drawings'to verify that the raised concrete pads are doweled to the concrete floors. l' Where plant drawings could not be used to verify the dowels exist, then other actions were
' taken. For example, an ultrasonic test was performed for the compressors discussed in Issue 9 ,
above. j The specific question above is for the raw water pumps. The reinforcement for the raised pads i is specifically shown in Section 11 of FPC Drawing 422-041. The "422" series of drawings !
are for the Auxiliary Building concrete floors, walls,.and equipment foundations. The raw l water pump sump is shown on drawings 422-001, 422-005, 422-007, 422-028, and 422-041.
Section 4 of GIP-2 allows for the use of existing plant drawings for determining anchorage ;
attributes. l t
1 ie roof of the s' mp consists of a reinforced concrete slab 3'-0" thick. The top of this slab is l at elevation 93'-0" (plant datum), and coincides with the top of the Auxiliary Building base l mat. On top of the base mat is the finished floor slab at plant elevation 95'-0" (nominal). The ;
pump foundation details on drawing 422-041 show a raised circular support pedestal with a !
3'-10" inside diameter, a thickness of l', and a finished height of 2'-31/2" above the base i mat. One inch of grout was then added to this. The finished floor was then poured around the
_ pedestal. Therefore, the top of the pedestal appears to be approximately 41/2" above the i floor (includes 1".of grout). The drawing shows the pedestal is reinforced with #5 bars in a !
circular pattern. The pedestal also is attached to the base mat with #5 dowels spaced radially l around the circumference of the pedestal. The pump is attached to a sole plate that is then l bolted to this raised pedestal with 1" diameter embedded bolts.
Based on a'through review of the plant drawings and the above information, FPC concludes !
there is adequate load transfer from the pumps'to the Auxiliary Building foundation system. i f
n l
e
^
I
U.S. Nuclear Regulatory Commission - Attachment D 3F1298-01 Page1of1 l l
1 ATTACHMENT D RESPONSE TO CABLE TRAY CALCULATION CONCERNS :
. 'Thej NRC Staff discussed - specific concerns related to suspended ladder-type (trapeze) calculations in Enclosure 2 of Reference 1. Those concerns were stated as follows:
The calculations do not contain a comprehensive discussion of the allowable stresses i used, including the basisfor their use and reference to the industry code criteria. We
- ' requested this information' during our audit. . Additionally, it appears that the i calculations have not been independently reviewed and venfied..
i Assumptions uired for obtaining the moment at the ceiling attachment point do not appear to be valid. It appears that the loadfrom the cables, which in certain cases are as much as 22 feet from the support, were transferred to 5 feet assuming rigid body load transfer and guided cantilever deformationfor the vertical member. Based on our ,
conservative calculations, we estimate that the stresses at the vertical member would be [
sigmjicantly higher. l RESPONSE. i As used above, " calculations" is a reference to the typical support calculations for cable trays, developed by the plant Architect / Engineer (A/E) as part of the original design of
- Crystal River Unit 3 (CR-3). All of the typical support calculations are contained in one calculation package. FPC agrees _ that the typical support calculations package is not
'l
. adequate since the design outputs in the calculation package are not verifiable.
Calculations prepared during original plant design and construction, in the early-1970s time period, did not always have an integral verification review as part of the calculation
. package. Separate design verification reviews were common, and were allowed by the
' A/E's quality assurance program. Efforts to locate the design verification package for this particular calculation package have been unsuccessful. Consequently, FPC will prepare a new calculation package, in accordance with current design and licensing basis ,
requirements, to demonstrate the adequacy of the typical supports for cable trays. This i
effort will be completed by December 15,1999, with other USI A-46 actions. I
, Additional actions to verify the seismic adequacy of raceway supports are discussed in Item i 11 in Attachment B of thi:: letter. !
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.i U.S. Nuclear Regulatory Commission _ Attachment E !
3F1298-01 Page1 of1 ATTACHMENT E .
REFERENCES
- 1. NRC to FPC letter .3N0498-07, dated April 10,'1998, " Audit Report of Unresolved ;
Safety Issue A-46, Seismic Implementation and Subsequent Evaluations of Related Issues at Crystal River Unit 3 (TAC No. M69440)"
- 2. NRC to FPC letter 3N0494-07, dated April 12,1994, " Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operatiag Reactors, Unresolved Safety Issue ,
(USI) A-46, Generic Letter (GL) 87-02, Crystal River Unit 3 (TAC No. M69440)"
- 3. NRC to FPC letter 3N0596-04, dated May 2,1996, " Crystal River Unit 3 - Evaluation of FPC's Plant Specific Criteria and Procedures for Implementing the Resolution of USI A-46 (Generic Letter 87-02) at Crystal River Unit 3" ,
- 4. FPC to NRC letter 3F1297-24, dated December 16,1997, " Supplemental Response for the Resolution of Unresolved Safety Issue A-46 (TAC No. M69440)" ,
- 5. . NRC to FPC letter 3N0197-20, dated January 28,1997, " Crystal River Unit 3 - Request for Additional Information on the Resolution of Unresolved Safety Issue A-46 (Generic Letter 87-02) (TAC No. M69440)"
. 6. FPC to NRC letter 3F0198-42, dated January 30,1998, " Supplemental Response for the
- Resolution of Unresolved Safety Issue (USI) A-46 (Generic Letter 87-02)
- 7. FPC to NRC letter 3F1295-18, dated December 31,1995, " Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, USI A-46, Generic Letter 87-02"
, 8. FPC Analysis / Calculation S97-0541,- Revision 0, " Confirmatory Study of Selected Safe Shutdown Equipment Anchorage for USI A-46" t
- 9. FPC to NRC letter 3F0397-28, dated March 27,1997, " Verification of Seismic Adequacy i of Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46" d
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