ML20198B398

Forwards RAI Re Licensee 960628 Submittal of Plant Specific Summary Rept in Response to USI A-46 Program at Plant

ML20198B398
Person / Time
Site:	Calvert Cliffs
Issue date:	12/19/1997
From:	Dromerick A NRC (Affiliation Not Assigned)
To:	Cruse C BALTIMORE GAS & ELECTRIC CO.
References
REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR TAC-M69435, TAC-M69436, NUDOCS 9801060303
Download: ML20198B398 (8)

Text

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C Deceraber 19,:1997 Mr. Charles H. Cruse -

Vice President - Nuclear Energy Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant -

1650 Calvert Cliffs Parkway Lusby, Maryland 20657.

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION ON THE RESOLUTION OF

/- UNRESOLVED SAFETY ISSUE A-46, CALVERT CLIFFG NUCLEAR POWER PLANT, UNIT NOS.1 AND 2 (TAC NOS. M69435 AND M69436)

By letter dated June 28,1996, the Baltimore Gas and Electric Company (the licensee) forwarded the plant specific summary report in response to the Unresolved Safety issue (USI) A-46 program at the Calvert Cliffs Nuclear Power _ Plant, Units 1 and 2 (CCNPP). The NRC has reviewed the summary report and determined that additionalinformation is needed to complete the review of the licensee's USI A-46 program, The information required is addressed in the enclosure, The staff requests that the additionalinformatien be provided within 90 days of receipt of this letter.

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If you have any questions regarding this letter, please contact me at (301) 415-3473, Sincerely, Original Signed by:

Alexander W. Dromerick, Project Manager Project Directorate I-1 Division of Reactor Projects -IIII Office of Nuclear R68etor Regulation Docket Nos. 50-317 and 50-318

Enclosure:

Request for Additional

, Information cc w/ encl: See next page DISTRIBUTION:

LIC . D merick RDI-1 R/F OGC B. Boger ACRS e I l

S, Bajwu . C. Hehl, Region I

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- DOCUMENT NAME: G:\CC12\M69435.RAI To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attacament/ enclosure "N"= No copy

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1. WASHINGTON, D.C. 30ees-4001 '

.,t Decenber 19, 1997

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Mr. Charles H. Cruse Vice President - Nuclear Energy Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, MD 20657

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION ON THE RESOLUTION OF UNRESOLVED SAFETY ISSUE A-46, CALVERT CLIFFS NUCLEAR POWER -

PLANT, UNIT NOS.1 AND 2 (TAC NOS. M69435 AND M69436)

By letter dated June 28,1996, the Baltimore Gas and Electric Company (the licensee) forwarded the plant cpecific summary report in response to the Unresolved Safety issue (USI) A-46 program at the Calvert Cliffs Nuclear Power Plant, Units 1 and 2 (CCNPP). The NRC has reviewed the summary report and determined that additionalinformation is needed to complete the review of the licensee's USl A~46 program. The information required is addressed in the enclosure.

The staff requests that the additionalinformation be provided within 90 days of receipt of this letter, if you have any questions regarding this letter, please contact me at (301) 415-3473.

Sincerely,

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.- cWr6 w der W.' Dromenck, Project Manag(er Alex Project Directorate 1-1 Division of Reactor Projects -I/II Office of Nuclear Reactor Regulation Docket Nos. 50-317 and 50-318

Enclosure:

Request for Additional Information cc w/ encl: See next page l  !

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i Mr. Charles Hi Cruse Calvert Cliffs Nuclear Power Plant Baltimore Gas & Electric Company Unit Nos.1 and 2 cc:

President Mr. Joseph H. Walter, Chief Engineer Calvert County Board of Public Service Commission of Commissioners Maryland 175 Main Street Engineering Division Prir.ce Frederick, MD 20078 6 St. Paul Centre Baltimore, MD 21202-6806 James P. Benriett. Esquire Counsel Kristen A. Burger, Esquire Baltimore Gas and Electric Company Maryland People's Counsel P.O. Box 1475 6 St. Paul Centre Baltimore, MD 21203 - Suite 2102 Baltimore, MD 21202 1631 Jay E. Silberg, Esquire -

Shaw, Pittman, Potts, and Trowbridge Patricia T. Birnie, Esquire 2300 N Street, NW Co-Director Washington, DC 20037 Maryland Safe Energy Coalition P.O. Box 33111 Mr. Thomas N. Pritchett, Director Baltimore, MD 21218 NRM Calvert Cliffs Nuclear Power Plant Mr. Loren F. Donatelt 1650 Calvert Cliffs Parkway NRr. Technical Training Center Lusby, MD 20657-4702 5700 Brainerd Road Chattanooga, TN 37411-4017 Resident inspector U.S. Nuclear Regulatory Commission

- P.O. Box 287 St. Leonard, MD 20685 Mr. Richard I. McLean, Manager Nuclear Programs i Power Plant Research Program Maryland Dept. of Natural Resources Tawes State Office Building, B3 Annapolis, MD 21401 Regional Administrator, Region I

( U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 l  !

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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NOS.1 AND 2 UNRESOLVED SAFETY ISSUE A-46

Reference:

Letter (and enclosures) from Baltimore Gas and Electric Company to USNRC,

" Summary Report for Resolution of USl A-46," June 28,1996

1. In Section 5.1.1 of Attachment 1 to your letter, the following effective grades are indicated:

Containments Elev.10' Intake structure Elev. 3' All other buildings Elev 45' In generating the " realistic median centered"in-structure response spectra in Appendix H to Attachment i for outliers resolution, you used NUREGICR-0098,84 percentile non-exceedance spectral shape and the realistic damping values recommended therein. You used 10% frequency peak shifting and indicated that the free field motion at the base mat elevation may be less than 60% of the ground level free field motion. The damping values mentioned are (1) 5% equipment damping in Section 1.1, (2) 7% structural damping for both steels and concrete elements in Section 2.0, and (3) the best estimate, lower and upper bounds of effective soil hysterstic damping of 0.08,0.10 and 0.06 in Section 3.0 of Appendix H.

On the basis of the information summarized above, we request the following additional information:

a) Since the NUREG/CR-0098 spectrum does not reflect the soil amplification at your site, what is the basis for using these spectra at the ground level for deconvolution to the foundation level of each structure? By doing this deconvolution, you are inappropriately reducing the spectra for the respense analysis. This is contrary to the method described in your submittal of September 18,1992, which was reviewed and accepted for use in resolving the A-46 issues. Provide the basis for your rev::.sd approach which deviates from your previous commitment.

b) is there any pro Jnd motion generated at the foundation level through deconvolution less than 60% of inat at the ground level? Provide your justification if such a condition exists.

- c) In all the figures provided,' 5% damping is indicated. Where is the 7% damping used?

d) The last sentence of Section 4 of Appendix H states, "For the Reactor Building the fundamental SSI mode frequency is shifted to the left, outside of the frequency range of maximum earthquake power resulting in the structural response being considerably reduced." The in-structure spectra with such shift are shown in Figures A1.12 and Enclosure j

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A1.13 in Appendix H. The staff bMieves that a single time history has been used. In accordance .

with SRP Section 3.7.1.1.1.b. Option 1. N Single Time History is used, N is required to satisfy a -  !

tat get power spectral density in addition to the design response spectra enveloping requirement. t Indicate if this requirement is met. K so, provide the documentation.-  !

2. In your reference letter, you stated that Baltimore Gas and Electric (BGE) committed to  !

= implement the Generic impbmontation Procedure, Revision 2 (GIP 2). You also stated that no significant or programrnatic deviation from the GlP 2 guidance were made during the USl ,

A 46 resolution process. Provide the woist-case items which deviate from the GlP 2 guideline l but were considered to be insignificant for Calvert Cliffs and also provide the bases for  !

categorizing them as such.

3. Referring to Tablea

• 21 and 5.2 2 of the attachment," Intent but Not Letter Cavest ,

Summary," for Units 1 and 2 respectively, each of the items listed in the table involve some  ;

_ degree of judgment or estimation by licensee personnel in concluding that the intent of

. applicable caverts were met. Provide the technical bases for the judgment or estimation that were md to determine the seismic adequacy for certain items (e.g.,1MS3g86,1C67, .

-- 1NB411, GPO542g,1CV517) listed in the tables, but not identified below. Also, provide the j following additionalinformation:  !

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.'s. For valve 2MOV50g, the top of the valve opeistor touches pipe insulation on the same I

- pip'ng, and the bleed off line touches support framing. You indicated that there is no j impact concem because the valve and piping will move together, and the piping has a nearby restraint. Provide additionalinformation to demonstrate that the vsive and tne piping, and the bleed off line and the support frame would not respond out of phase to one another and impact each other during an earthquake.  ;

b. Valves 1MOV501,1MOV514,1MOV504,2CV517, etc. are independently supported frem i the connected piping. Provide an example calculation in support of yourjudgment to demonstrate hat the specific piping and valve support configuration would not cause an i overstressed wondition, f

c. Valves 1PCVd510,1PCV4520, etc., are mounted on piping less than 1"in diameter. I Provide the analysis, in acco1ance with the GIP 2, to ensure that the structural integrity of the valves and the connected piping is satisfied. ]

L 4. Referring to Tables 5.3-1 and 5.3 2,

• Equipment Outiler Description and Resolution Summary," for Units 1 and 2 respouuvely, we find the analysis method employed for .i resolution of each of the items listed in the table to be unclear Provide a more detailed  !

description of the analyr's . employed for resolution of r representative sample outliers and ,

the basis for the juch. . ant used.- Also, provide the following additionalinformation:

a, Valves 1CV5210,1CV5150,1CV5152,1CV5212,1CV5153,1CV5155, etc. were klontified I to be outliers because ti eir valve bodies werg manufactured of cast iron. Provide an - 6 analysis to demonstrate that these valves will not be overstressed.  ;

c. Panel iPYS427A was identified to be an outlier because the natural frequency is within the l range where the required response spectra are greater that 1.5 times the bounding I

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spectrunt. You proposed that this outlier be resolved by plant modification that eliminates l the need for the component Provide an explanation as to how the panel function is no j longer needed.

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c. There are hems (e.g., SwHohes 1PS5431,1P45432,1PS$433,1PS5434, etc.) for which both the ou' lier description and the odlier resolution address the same Rule-of the-box ,

number. Please explain in detail using any of the switches as an example.

5. In Section 6 of Attachment 1 to your letter, outlier items for tanks and heat exchangers are l Identified and listed in Table d.1, All of the outliers were resolved by additional or further ,

analysis except for the condensate storage tanks and refuel.ng storage tanks. For these -

tanks, the calculated seismic overtuming moment was 8161 kip ft which exceeds the .,

overtumine capacity of 6292 kip-ft. A dynamic analysis accounting for soil-structure interaction was performed by Stevens & Associates using its developed computer code,

- SUPER SASSI/PC. The results showed overtuming moments in the magnitude of 5395 kip ft and 6000 kip ft for shear wave velocities of 760 fps and 1600 fps, respectively. However, it '

is nur understanding that the staff has not previously reviewed and approved the SUPER SASSI/PC code. Provide the validation documents which should contain sufficient '

information for the staffs review, especially in the areas of application at Calvert Cliffs for resolution of the USI A 46 issues. Also, provide for each type of outlier a sample calculation  :

to show how the outlier was resolved.

6. In reference to Section 6.2 of the attachment, " Summary of Outliers" for Tanks and Heat Exchangers, you indicate that an expansion anchor capacity reduction factor (CRF) of 0.75 rather than 0.5 for unknown bolt type anchors was used in accordance with Revision 3 of the GlP (GlP 3), dated July 31,1995. On the basis of its Supplemental Safety Report No. 3 (SSER No. 3) dated December 4,1997, on the review of the updated GIP 3 dated May 16,  ;

1997, the staff did not accept this CRF. Therefore, the staff request that you re-evaluate l eq@ ment types for which the unknown expansion anchor CRF of 0.75 was used in evaluating their seismic adequacy. The evaluation should be in compliance with SSER No. 3 and the updated GIP 3. Also, provide the results for the staffs review.

7. The NRC staff has concems about the way the USl A 46 cable trays and conduit raceways issue was being disposed of by some USl A-46 licensees. The staff issued requests for i additional information (RAls) to several licensees on this issue. SQUG responded instead of the licensees because SQUG considered the RAls to be generic in nature. The staff issued a subsequent RAI to SQUG as a follow up to their response. However, the staff found that the correspondete s with SQUG did not achieve the intended results, in that it did not address the identified technical concems of the staff. Therefore, we are stating our concems in the l following discusion. ,

The GIP procedure recommended performing what is called "a limited analytic evaluation" for

'selectad cable and conduit raceway Tuppor'.s. The procedure further recommendect that when -l a certain cable tray system can be ju;ged to be ductile and if the verticalload capacity of the

anchorage can be established by a load check using three time 6 the dead weight, no further evaluation is needed to demonstrate lateral resistance to vibration from earthquakes. The _i staff has concems with the manner in which these simplified GlP criteria were implemtnted at .

your pis,nt, 1

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The GIP procedure eliminates horizontal force evaluations by invoking ductility. However, some so-called non ductile cable tray support systems would eventually become ductile by inelastic deformation, buck!:ng or failure of the non-ductile cable tray supports and ,

members. This pmcedure is a basic departure from conventional methods of engineering i evaluation and the GlP does not provide an adequate basis for dealing with those cable trays that are initially judged to be non-ductile but are eventually called ductile by postulating failure of the lateral supports. If this procedure was followed for eliminating cable trays from further assessment at your plant, then all the cable trays could conceivably be screened out from the A-46 evaluatiers We request that you provide the following information to enable our assessment and safety evaluation of cable trays at your plant, l

n. Define ductility in engineering terms as used at Calvert Cliffs for the USl A-46 evaluation.

Clarify how this definition is applied to actual system configurations at Calvert Cliffs plant i consistently for the purpose of analytical evaluation. l l

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 horizontalload evaluation. Indicate the approximate percentage of such raceways au compared with the o*pulation 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 the ductile raceways need not to be evaluated for horizontalload.

When a reference is provided, state the page aumber and paragraph. The reference should be self contained, and not refer to another reference.

e. In the evaluation of the cable trays and raceways, if the ductility of the attachments is assumed in one horizontal direedon, does it necessarily follow that the same system is ductile in the perpendicular direction? If yes, provide the basis of this conclusion, if it is not ductile in the perpendicular direction, how was the seismic adequacy of the attachments evalu/ed?

f. Discuss Lny raceways and cabl6 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.

Pro ride 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. Discuss how will they 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 taetway), including the configurations (dimension, member size, supports, etc.).

8. In Section 8 of the attachment, you stated that there are 35 unresolved outliers and the schedule for resolving them is provided in Sectio'. 5.3 of the enclosure to your letter.

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+6a However, there is no resolution schedule provided in Section 5.3. On page 2 of your letter, -- l you indicated that you plan to resolve all outliers by July 1998. Explain the safety implications j for not resolving these outliers, in accordance with item 17 in Section g.1 of the Gip 2. You  :

are also requested to elaborate en your decision to deler the resolution of identitled outliers 1

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and your evaluation in support of the conclusion that the operebWty of afteMed pierd  ;

equipmerd will not be effected by your decision.  ;

g. In Appendix G cf the attachmerd, the third party audit report and responses indicates that the i instrument and control panel (1NB130) is attached to a block and was not judged to be i acceptable by BGE, Clarify why this equipmerd is listed as line no. 6053 in the Calvert Cliffs  !

Unit i Safe Shutdown Equipment List (SSEL) but not documented in the 8creening l Veritication Data Sheet (8VDS)in Appendix D. Also, provide the status and resolution j regarding the seismic adequacy of this control panel. J

10. Referring to the in structure fssponse spectra provided in your 120-day-response to the  !

NRC's request in Supplement No.1 to Generic Letter 87-02, dated May 22,1992, provide the l following information. l

a. Identify structure (s) which have in structure response spectra (5% critical damping) for j elevations with,,i 40-feet above the effective grade, which are higher in amplitude than 1.5  ;

times the SQUG Bounding Spectrum. i i

b. With respect to the comparison of equipment se rmic capacity and seismic demand,  !

Indicate which method in Table 4-1 of GIP 2 v<as 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 GlP 2, provide a technical i justification for not using the in structure resp >nse spectra provided in your 120-day.  !

response, it appears that some A 46 licensees are making an incorrect comparison j between their plant's safe shutdown earthquake (SSE) ground motion response spectrum  ;

and the SQUG Bounding Spectrum. The SSE ground motion response spectrum for j most nuclear power plants is defined at the' plant foundation level. The 80U0 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 i amplitudes at the foundation level and those at the ground surface. However, for sites where a structure is founded on sh w soil, the amplification of the ground motion from

, the foundation level to the ground surface may be significant.  :

- c. For the structure (s) identified in item (a) above, provide the in-structure response spectra  !

' designated accordog to the height above the effective grade. If the in-structure response -

spectra identified in the 120 day-response to supplement No.1 to Generic Letter 87-02 -!

were not used, provide the response spectra that were actually used to verify the seismic -

adeqyacy of equipment within the structures identified in item (a) above. Also, provide a

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comparison'of these spectra to 1.5 times the Bounding Spectrum.  !

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