ML17334A636
| ML17334A636 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/23/1998 |
| From: | John Hickman NRC (Affiliation Not Assigned) |
| To: | Fitzpatrick E INDIANA MICHIGAN POWER CO. |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR GL-87-02, GL-87-2, TAC-M69437, TAC-M69438, NUDOCS 9801290140 | |
| Download: ML17334A636 (11) | |
Text
January 23, 1998 Mr. E. E. Fitzpatrick, Vice President Indiana, Michigan Power Company Nuclear Generation Group 500 Circle Drive Buchanan, Ml 49107
SUBJECT:
SECOND REQUEST FOR ADDITIONALINFORMATIONON THE RESOLUTION OF UNRESOLVED SAFETY ISSUE (USI) A-46, D.C. COOK NUCLEAR PLANT, UNIT NOS.,1 AND 2 (TAC NOS. M69437 AND M69438)
Dear Mr. Fitzpatrick:
In your letter dated January 30, 1996, you provided the plant-specific summary report in accordance with commitments relating to Generic Letter 87-02 on the resolution of unresolved
, safety issue (USI) A-46 program at the D. C. Cook Nuclear Plant, Unit 1 & 2. On March 10, 1997, you transmitted a response to the staffs RAI transmitted on October 21, 1996. The staff has reviewed the response, and determined that additional information is needed to complete our reviews. Attached is our second RAI.
Ifyou have any questions regarding this matter, please contact me at (301) 415-3017.
Original signed by:
Docket Nos. 50-315 and 50-316
Enclosure:
As stated cc w/encl: See next page DIST BU 0 we c:
Docket File PUBLIC PD3-3 R/F SHou RSavio John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation DOCUMENT NAME: G:iDCCOOKic069437.RAI To receive a co of this document. indicate In the box: "C"= Co w/o attachment; E"-"Copy w/attachment; "N"-"No Copy OFFICE PD3-3:PM C
NAME JHickman PD3-3:LA C
EBarnhill DATE
/ /g9/98 f /g /98 OFFICIAL RECORD COPY
'P801290L40 980123 PDR ADOCK 05000315 P
PDR IffyBK'f'lEWI CSPIt flllfflliffllifllllilffiifllflilffilfill'
II 1
1t
January 23, 1998 Mr. E. E. Fitzpatrick, Vice President Indiana Michigan Power Company Nuclear Generation Group 500 Circle Drive Buchanan, Ml 49107
SUBJECT:
SECOND REQUEST FOR ADDITIONALINFORMATIONON THE RESOLUTION OF UNRESOLVED SAFETY ISSUE (USI) A-46, D.C. COOK NUCLEAR PLANT, UNIT NOS.
1 AND 2 (TAC NOS. M69437 AND M69438)
Dear Mr. Fitzpatrick:
In your letter dated January 30, 1996, you provided the plant-specific summary report in accordance with commitments relating to Generic Letter 87-02 on the resolution of unresolved safety issue (USI) A-46 program at the D. C. Cook Nuclear Plant, Unit 1 8 2. On March 10, 1997, you transmitted a response to the staffs RAI transmitted on October 21, 1996. The staff has reviewed the response, and determined that additional information is needed to complete our reviews. Attached is our second RAI.
Ifyou have any questions regarding this matter, please contact me at (301) 415-3017.
Original signed by:
Docket Nos. 50-315 and 50-316
Enclosure:
As stated cc w/encl: See next page S
U
/e Docket File PUBLIC PD3-3 R/F SHou RSavio John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation DOCUMENT NAME: G:)DCCOOKttCO69437.RAI To receh/e a co of this document. indicate in the box: "C = Co w/o attachment; "E" ~ Copy w/attachment; N ~ No Copy OFFICE DATE PD3-3:PM C
PD3-3:LA C
JHickman
/ /g9/98 EBarnhill f /g /98 OFFICIAL RECORD COPY
gp,S REGII
+4 'o Op I
ClO IVl0 V/
\\ 0
++*++
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&0001 January 23, 1998 Mr. E. E. Fitzpatrick, Vice President Indiana Michigan Power Company Nuclear Generation Group 500 Circle Drive Buchanan, Ml 49107
SUBJECT:
SECOND REQUEST FOR ADDITIONALINFORMATIONON THE RESOLUTION OF UNRESOLVED SAFETY ISSUE (USI) A-46, D.C. COOK NUCLEAR PLANT, UNITNOS. 1 AND 2 (TAC NOS. M69437 AND M69438)
Dear Mr. Fitzpatrick:
In your letter dated January 30, 1996, you provided the plant-specific summary report in accordance with commitments relating to Generic Letter 87-02 on the resolution of unresolved safety issue (USI) A-46 program at the D. C. Cook Nuclear Plant, Unit 1 8 2. On March 10, 1997, you transmitted a response to the staffs RAI transmitted on October 21, 1996. The staff has reviewed the response, and determined that additional information is needed to complete our reviews. Attached is our second RAI.
Ifyou have any questions regarding this matter, please contact me at (301) 415-3017.
Docket Nos. 50-3.15 and 50-316
Enclosure:
As stated cc w/encl: See next page John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Indiana Michigan Power Company Donald C.
Cook Nuclear Plant Units 1 and 2
CC:
Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Attorney General Department of Attorney General 525 West Ottawa Street
- Lansing, MI 48913 Township Supervisor Lake Township Hall P.O.
Box 818
- Bridgman, HI 49106 Al Blind, Site Vice President Donald C.
Cook Nuclear Plant 1 Cook Place
- Bridgman, HI 49106 U.S. Nuclear Regulatory Commission Resident Inspector's Office 7700 Red Arrow Highway Stevensvi 1 1 e, HI 49127 Gerald Charnoff. Esquire
- Shaw, Pittman, Potts and Trowbridge 2300 N Street.
NW.
DC 20037 Mayor, City of Bridgman P.O.
Box 366
- Bridgman, MI 49106 Special Assistant to the Governor Room 1 - State Capitol
- Lansing, HI 48909 Drinking Water and Radiological Protection Division Michigan Department of Environmental Quality 3423 N. Hartin Luther King Jr Blvd P:0.
Box 30630, CPH Mai lroom Lansing.
HI 48909-8130 Steve J.
Brewer Indiana Michigan Power Company Nuclear Generation Group 500 Circle Drive
'uchanan, MI 49107 E.E. Fitzpatrick, Vice President Indiana Michigan Power Company Nuclear Generation Group 500 Circle Drive
- Buchanan, MI 49107
SECOND REQUEST FOR ADDITIONALINFORMATION ON, RESOLUTION OF USI AA6 ON EQUIPMENT SEISMIC ADEQUACY AT D. C. COOK NUCLEAR PLANT, UNIT 1 & 2 DOCKET NOS 50-315 AND 50-316
Reference:
1.
Letter (AEP:NRC:1040C) from Indiana Michigan Power Company to NRC with three attachments, dated January 30, 1996.
Letter (AEP:NRC:1040E) from Indiana Michigan Power Company'to NRC with six attachments, dated March 10, 1997.
In page 4 of the letter in Reference 1, the licensee indicates its intent to apply the USI A-46 methodology to future verification of seismic adequacy of the repair/replacement of equipment, including the scope of equipment identified as part of Regulatory Guide 1.97.
However, the staff position in Item 2 of Section I.2.3 of the SSER-2, which clarifies Section 2.3.3 of the GIP-2 regarding revision of plant licensing bases, is that any previous licensing commitments, such as for RG 1.97 and TMIAction Plan item II.F.2, are not superseded by the resolution methods of the GIP. Clarify your position regarding the means you intend to employ for incorporating the GIP-2 methodology into your licensing basis for verification of the seismic adequacy of new and replacement equipment.
As an example, also clarify your commitment with regard to the applicability of the A'-46 methodology to new and replacement Category 1 equipment included in the SSEL that are associated with RG 1.97 or TMIAction Plan item II.F.2.
Section 3.3 ofAttachment 2 (Reference
- 1) indicates that a peer review was performed which covered all seismic evaluation areas, including review of draft reports, sample walkdowns, and review of documentations.
Provide a summary of the peer review, including a description of major findings, recommendations, and the basis for the peer reviewer's conclusions, especially on USI AP6 program adequacy and verification of conformance to GIP-2 and SSER-2 guidelines in the licensee's screening walkdowns and seismic evaluation.
Is this an independent review action, or a joint action with the licensee's review team?
Describe your follow up actions as a result of this peer review.
3.
WI N
vd
-02 2
9 h
a.
Identify structure(s) which have in-structure response spectra (5% critical damping) for elevations within 40-feet above the effective grade, which are higher in amplitude than 1.5 times the SQUG Bounding Spectrum.
b.
With respect to the comparison of equipment seismic capacity and 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 ENCLOSURE
2 in the structure(s) identified in Item (a) above.
Ifyou 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 ground surface.
However, for sites where a structure is founded on shallow soil, the amplification of the ground motion from the foundation level to the ground surface may be significant.
For the structure(s) identified in Item (a) above, provide the in-structure response spectra designated according to the height above the effective grade.
Ifthe in-structure response spectra identified in the 120-day-response to Supplement No.
1 to Generic Letter 87-02 was not used, provide the response spectra that were actually used to verify the seismic adequacy of equipment within the structures identified in Item (a) above.
Also, provide a comparison of these spectra to 1.5 times the Bounding Spectrum.
In a letter dated May 31, 1996, the licensee submitted its commitment regarding the outlier resolution schedule.
The proposed time frame to complete all outlier resolution ranges from the end of 1996 to the end of 1999, depending on the determination whether a specific outlier will be resolved by analysis or by modification. As a number of safety-related components in the safe shutdown path have been identified as outliers, their seismic adequacy may be rendered questionable and their conformance to the licensing bases uncertain.
Elaborate on your plans for scheduling the resolution of identified outliers, and your evaluation in support of the conclusion that the licensing bases for the plant will not be affected by the outliers resolution schedule.
Describe the extent to which the seismic margin methodology, described in the report EPRI NP-6041, was used in the D. C. Cook a<6 program, including outlier resolutions for tanks and heat exchanges.
Since this methodology is known to yield analytical results which are not as conservative as what might be obtained by following the GIP-2 guidelines, it is generally not acceptable for the a-46 program.
Therefore, for each deviation from the GIP-2 guidelines, in situations where the margin methodology is utilized, identify the nature and the extent of the deviation, and provide the justification for its acceptance.
a note under Table 2-1 in Section 2.3 ofAttachment 2 (Reference
- 1) indicates that the damping values defined in GIP-2 were used for the USI a<6 effort, and that for the majority of the equipment classes, 5% damping was'used.
Identify the cases in which damping values higher than those specified in GIP-2 were used.
Provide the basis of using such damping values.
Items 44 and 45 in Table 4-5 ofAttachment 2 (Reference
- 1) indicate that batteries identified as 1-BATT-ABand 1-BALI-CDare just over 10-years old and were designated as outliers. The report further indicates that the outlier resolution was achieved in both cases by conducting evaluation to determine their seismic adequacy.
Provide details of how the seismic adequacy was verified for these batteries.
Item 1 in Table 4-7 ofAttachment 2 (Reference
- 1) indicates that a total of 101 equipment items were similar to but different from components in the class of 21 equipment, and that necessary evaluations were made to meet the intent of the GIP caveats.
Provide details regarding these evaluations and how the intent of caveats
~
were met. These details should include the following information in a tabular form for each of these equipment:
- a. Equipment description
- b. Caveat Number in the GIP-2
- c. Description of deviation from the GIP-2 caveat
- d. Justification for resolution In Appendix E of Attachment 3 (Reference 1), provide the basis for Footnotes (6) and (10) at the end of the tabulation of the Unit 1 essential relays'apability vs. demand summary, which indicate that these type of relays use only normally open contacts for essential functions, thus, increasing the seismic capability, and removing the low ruggedness restriction for these relays.
In addition, explain the distinction between the two footnotes.
In Appendix E of Attachment 3 (Reference 1), Footnote (9) of the Unit 1 essential relays'apability vs. demand summary table indicates that the amplification of certain relay panel was calculated using case specific analysis.
Considering relay panel under Tag Number 2-88X1-BCB as an example, provide an illustration how the case specific analysis was performed.
In Appendix E of Attachment 3 (Reference 1), Footnote (20) of the Unit 1 essential relays capability vs. demand summary table indicates that for relays located between elevations for which floor response spectra were developed, the next highest elevation was used.
As indicated in many cases, the developed response spectra may not pertain to the same building in which the relays to be seismically verified are located.
Explain how the pertinent response spectra were selected for this population of relays.
Provide the basis for Footnote (17) of the Unit 1 essential relays capability vs. demand summary table (Reference 1), which indicates that the floor response spectrum at elevation 591 ft. of the Turbine Building was used for relays located at elevation 594 ft.
of the Screenhouse.
Explain the inconsistency between this footnote and Footnote (20), which indicates that the spectrum at the higher elevation was used for the relays'valuation.
The NRC staff has concerns about the way the a<6 cable trays and conduit raceways issue was being disposed of by licensees.
We issued requests for additional information
(RAI) to several licensees on this issue.
SQUG responded instead of the licensees because SQUG considered the RAI to be generic in nature.
The staff issued a subsequent RAI to SQUG as a followup 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 concerns of the staff. Therefore, we are stating our concern 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 ifthe 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 concerns 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.
Ifthis 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 to the following items to elicit information that would support our safety evaluation of cable trays at your plant.
a)
Define ductility in engineering terms.
Clarify how this definition is applied to actual system configurations consistently for the purpose of analytical evaluation.
b)
Provide the total number of raceways that were selected for worst-case analytical calculations and were classified as ductile in your a-46 evaluation and for which you did not perform'a horizontal load evaluation.
Indicate the approximate percentage of such raceways as compared with the population selected for analytical review. Discuss how the ductility concept is used in your walkdown procedures.
c)
Describe the typical configurations of your ductile raceways (dimension, member size, supports, etc.)
d)
Justify the position that ductile raceways need not to be evaluated for horizontal load. When a reference is provided, state the page number and paragraph.
The reference should be self contained, and not refer to another reference.
e)
In the evaluation of the cable trays and raceways, ifthe ductility of the attachments is assumed in one horizontal direction, does it necessarily followthat the same system is ductile in the perpendicular direction?
f)
Discuss raceways and cable trays including supports in your plant that are outside of the experience data.
Explain what criteria are used for establishing their safety adequacy and specify your plan for resolution of outliers that did not meet the
acceptance criteria. Provide examples of the configurations of such raceways and cable trays including supports.
Also, indicate the percentage of cable trays and raceways outside the experience data in relation to the population of raceways and cable trays examined during the walkdowns of the safe shutdown path.
How are they going to be evaluated and disposed?
g)
Submit the evaluation and analysis results for four of the representative sample raceways (one single non-ductile, one single ductile, one multiple non-ductile, and one multiple ductile raceway), including the configurations (dimension, member size, supports, etc.).
14.
In your program were there any deviations from the GIP guidance?
Provide the worst-case items (from the safety point of view) which deviate from the GIP-2 guideline but were categorized as not being significant.
In addition, you are requested to submit the definition of "safety significant" that the walkdown crew used and provide a justification of why the definition is adequate.
15.
Indicate whether you found an anchor type (e.g., lead cinch anchor) not covered by the GIP-2 during the walkdown. Ifyes, how did you resolve the issue?