ML18039A426
| ML18039A426 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/13/1998 |
| From: | De Agazio A NRC (Affiliation Not Assigned) |
| To: | Scalice J TENNESSEE VALLEY AUTHORITY |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR GL-87-02, GL-87-2, TAC-M69431, TAC-M69432, NUDOCS 9807160173 | |
| Download: ML18039A426 (15) | |
Text
Mr. J. A. Scalice Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801 July 13, 1998
SUBJECT:
UNRESOLVED SAFETY ISSUE A-46 REQUEST FOR ADDITIONAL INFORMATION:BROWNS FERRY NUCLEAR PLANT, UNITS 2 AND 3 (TAC M69431 AND M69432)
Dear Mr. Scalice:
By letter dated June 28, 1996, the Tennessee Valley Authority (TVA)provided the plant-specific summary report in accordance with its commitment relating to Generic Letter 87-02 on the resolution of unresolved safety issue A-46 program at the Browns Ferry Nuclear Plant Units 2 and 3. The staff has continued its review of the summary report and has determined that additional information is necessary to complete the review of.TVA'sAQ6 submittals.
The enclosure identifies the additional information needed.
Please provide your response by August 31, 1998.
Sincerely, Original signed by Albert W. De Agazio, Sr. Project Manager Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-260, 50-296 Serial No. BFN-98-015
Enclosure:
Request for Additional Information V
cc w/enclosure:.See next page uv" DISTRIBUTION:
FHebdon Docket File PUBLIC ADeAgazio BFN r/f JZwolinski ACRS PYChen 'Clayton OGC DOCUMENT NAME: G:>BFN>69431RAI.42 To receive a copy of this document, indicate in the box:
"C" = Copy without attachment/enclosure "E" -"
Co y with attachment/enclosure "N" = No copy OFFICE PD I I -3/PN PDII-3/LA PDII-3/D NANE ADeAgazi BCla ton FHebdon DATE 07/
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Mr. J. A. Scalice Tennessee Valley Authority BROWNS FERRY NUCLEARPLANT CC:
Senior Vice President Nuclear Operations Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Jack A. Bailey, Vice President Engineering 8 Technical Services Tennessee. Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. C. M. Crane, Site Vice President Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609 General Counsel Tennessee Valley Authority ET 10H 400 West Summit Hill Drive Knoxville, TN 37902 Mr. Raul R. Baron, General Manager Nuclear Assurance Tennessee Valley Authority 5M Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Karl W. Singer, Plant Manager Browns Ferry Nuclear Plant Tennessee. Valley Authority P.O. Box 2000 Decatur, AL'5609 Mr. Mark J. Burzynski, Managar Nuclear Licensing Tennessee Valley Authority 4X Blue Ridge 1101 Market Street
, Chattanooga, TN 37402-2801 Mr. Timothy E. Abney, Manager Licensing and Industry Affairs Browns Ferry Nuclear Plant Tennessee Valley Authority P..O. Box 2000 Decatur, AL35609 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 61 Forsyth Street, SW., Suite 23T85 Atlanta, GA 30303-3415 Mr. Leonard D. Wert Senior Resident Inspector U.S. Nuclear Regulatory Commission Browns Ferry Nuclear Plant I0833 Shaw Road Athens, AL35611 State Health Officer Alabama Dept. of Public Health 434 Monroe Street Montgomery, AL36130-1701 Chairman Limestone County Commission 310 West Washington Street Athens, AL 35611
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BROWNS FERRY NUCLEAR PLANT, UNITS 2 & 3 DOCKET NUMBERS 50-260 and 50-296 REQUEST FOR ADDITIONALINFORMATION REGARDING THE VERIFICATIONOF SEISMIC ADEQUACY OF MECHANICALAND ELECTRICALEQUIPMENT Referring to the in-structure response spectra (ISRS) discussed in your response (Reference 1) to the Nuclear Regulatory Commission (NRC's) request in Supplement No. 1 to Generic Letter 87-02, dated May 22, 1992, the following information is requested:
(a)
Identify structure(s) which have ISRS (5% critical damping) for elevations within 40-feet above the effective grade, which are higher in amplitude than 1.5 times the Seismic Qualification Utilities Group 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 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 ISRS provided in Reference 1.
(c)
For the structure(s) identified in Item (a) above, provide the ISRS designated according to the height above the effective grade.
Ifthe ISRS identified in Reference 1 were not used, provide the response spectra that were 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.
(d)
Identify all the safe shutdown equipment list (SSEL) equipment installed in the Diesel Generator Buildings'identified in Item (a) above, and provide information in a tabular form with the pertinent information similar to those in Appendix D of Enclosure 1 and Appendix A of Enclosure 2 to Reference 2, for the verification of seismic adequacy of each equipment identified, in view of the concerns identified in Items (b) and (c) above.
In Section 5.1.3 of Enclosure 1 to Reference 2, you indicated that bounding calculations were performed to address the seismic adequacy of the anchorage.
State whether you followed the guidelines provided in Appendix C of the GIP-2 procedure in determining nominal allowable capacities and capacity reduction factors.
Submit sample ENCLOSURE
r I
calculations for each of the following anchor types: expansion anchor, welded anchor, cast-in-place bolt and headed studs, cast-in-place J-bolt; and grouted-in-place bolt, using the worst-case bounding anchorage evaluations for various equipment classes (MCC, switchgear, transformers, distribution panels, battery chargers, electrical cabinets and mechanical equipment, etc.).
Indicate whether an anchor type, e.g., lead cinch anchor, not covered by the GIP-2 was used for SSEL equipment anchorage at Browns Ferry Nuclear Plant (BFN) Units'2 8 3 during the Unresolved Safety:Issue (USI) AC6 walkdown. Ifyes, how did you resolve the issue?.
You indicated in Appendix E of Enclosure 1 to Reference 2 that an anchorage evaluation was performed for equipment (CRD/hydraulic control unit, I.D. Nos. 2-HCU-85,1-185 and 3-HCU-85,1-185), which was not specifically addressed by the GIP-2.
Provide detailed information (equipment dimension, anchorage type and dimension, etc.) with your evaluation for the verification of seismic adequacy of this equipment and anchorage.
In Section 6.1.1 of Enclosure 1 to Reference 2, you stated that no large vertical flat-bottom tanks are on the BFN Seismic Review SSEL. Since USI A-40 is a part of USI A-46 for BFN, clarify whether this statement is still true. Ifnot, provide the resolution of these large vertical flat-bottom tanks for closure of USI ARO. Indicate whether you used the SMA methodology described in the Electric Power Research Institute NP-6041 report for the resolution of tanks and heat exchangers.
The staff has noted that the SMA methodology is known to yield analytical results which may not be as conservative as those obtained by following the GIP-2 guidelines, hence, it is generally not acceptable for the USI A46 program.
Describe the extent to which the method was used in your USI A-46 program.
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 a technical justification for its acceptance.
You indicated in Section 6.1.2 and Appendix H-2 of Enclosure 1 to Reference 2 that a heat exchanger (Equipment I.D. No. 3-HEX-74-900D) was not covered by the GIP-2 procedure and was classified as an outlier. Provide, for staff review, the calculation (Gale. 50147-C004) for the resolution of the outlier. Provide also the information concerning the seismic adequacy of the following tanks, including the Screening Evaluation Work Sheets (SEWS) for each item: (1) CAD/Nitrogen tanks, (2) DG 7-day fuel oil tanks, and (3) diesel generator starting air receivers.
The GIP-2 procedure recommended that the licensee perform a limited analytical evaluation for selected raceways and cable trays. The procedure 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
'ead weight, no further evaluation is needed to demonstrate lateral resistance to vibration from earthquakes.
l a)
Provide descriptions of typical configurations ofyour ductile raceways (dimension, member size, supports, etc.).
b)
Discuss the configuration of raceways and cable trays that are outside of the experience data, and provide an estimate of the percentage ofthese installations with respect to the whole population of raceways.
Discuss your approach for the evaluation and disposition of these installations.
8.
Section 7.3 and Appendix l-2 of Enclosure 1 to Reference 2 presents cable tray and conduit raceway outliers and their resolutions.
Provide calculations for the following outliers: ¹1 9-02, ¹44-01, ¹22-03 and ¹35-01.
9.
Ifthermal-lag panels are attached to a cable tray system, discuss how the changes in weight have been incorporated in the GIP evaluation of these systems and their supports.
10.
In Appendix C of Enclosure 1 to Reference 2, there are equipment items designated as "ROB" class, which were seismically verified using the "rule of box" as outlined in GIP-2. Identify which "Box" each of the following equipment items belongs to and provide, for staff review, the evaluation performed to determine the seismic adequacy of each item, including mounting and/or anchorage evaluations:
D Rmrll;i~
C.
1018 1018 3053 3065 9168 2-FT-74-50 2-BKR-402 2-LT-3-58B 2-NM-92-7/41A 2-PX-64-67B RHR LOOP I Flow Transmitter RHR LOOP I Flow Indicator Breaker RPV Level Transmitter Channel "A" IRM Indicator Power Supply (PNL 2-9-19; Supports 2-PL-64-67B) 9393 2-AMP-092-0007/41A IRM CH. "A"Voltage Preamplifier 7-34A 33055 3-XR-64-159 Torus and Drywell Pressure Instrument 11.
Appendix F-1 of Enclosure 1 to Reference 2 provides a list of instances where a special exception to enveloping the seismic demand spectrum is used, i.e., the seismic capacity spectrum of an equipment envelops the seismic demand spectrum only at, and above, the conservatively-estimated lowest natural frequency of the equipment.
In Section II.4.2.1 of the staffs Supplemental Safety Evaluation Report No. 2 (SSER-2) dated May 22, 1992, the staff provided some cautions with regard to the use of this
exception.
Submit, for each of the following sample equipment items, (1) descriptions of the equipment including, at the least, the dimensions, internal components, mounting or anchorage conditions, and special features, (2) comparison of seismic capacity spectrum with the seismic demand spectrum, (3) natural frequencies of the equipment including equipment assembly, subassemblies, door panels, internal structures and components as applicable, and how they were estimated, (4) justification of the adequacy of partial enveloping of the demand spectra, and (5) SEWS sheets:
b.
C.
SEEL~
9028 9020 9014
'006 9285 9406 9305 9040 ID 2-BDBB-268-002E 2-BDBB-231-0002A 0-BDAA-211-000C 2-XFA-253-0002A1 2-JBOX-268-5991 2-CHGD-283-A1-2 2-LPNL-925-247A 2-PNLA-009-003A
~D~~cr'o 480 RMOV BD 2E 480 SHDN BD 2A 4KVSHDNBDC 480V-120/280V XFMR FOR l8C BUS 2A MG SET 2DA CONTROL STATION (2-HS-268-0002DA) 24V NEUTRON BATTERY CHARGERS A1-2 LOCALPANEL 2-25-247A (CAD DRYWELL8 SUPP.
CHAM. V.)
PANEL 9-3A 39117 3-PNLA-009-0005 REACTOR CONTROL PANEL 39216 3-LPNL-925-655A DIV1 LOAD SHED LOGIC PANEL 12.
In Appendix F-2 of Enclosure 1 to Reference 2, SSEL Item Nos. 9119, 9122, 9125, and 9128 (250 V Battery SB-A, B, C, and D) are batteries with multi-tiered racks.
You concluded that these batteries on racks met the intent of GIP-2 caveat GR4 for batteries on racks even though the GIP-2 caveat limits the applicability of Generic Equipment Ruggedness Spectra only for batteries supported on two-step or single-tiered racks with longitudinal cross-braces.
The items should have been identified as outliers and
, resolved accordingly. The footnote indicated that these racks are represented in a portion of the seismic experience database.
Provide, for staff review, the documentation of this seismic experience database, and justify the seismic adequacy of these battery racks and the batteries at BFN, taking into account the potential amplification through the racks to the center of gravity of the batteries.
Also submit the SEWS for these items.
13.
Appendix G of Enclosure 1 to Reference 2 provides a summary of outliers and resolution methods for mechanical and electrical equipment.
Submit, for staff review, the detailed resolution documentation including SEWS sheets for the following sample items:
NISEI~
UI b.
39008 9160 1004 3001 39202 39204 3-BDBB-268-003A 0-XFA-082-000AA 2-PMP-74-5 2-FCV-1-14 3-PNLA-082-0003C 3-PNLA-925-0031 480 V RMOV BOARD 3A DG-A Neutral Gm XFMR
, RHPJPump 2A MSIV"A" Inboard Iso Valve DG 3C Elect. Control Cabinet Local Panel 3-25-31 14.
In your response (Reference 3) dated January 19, 1993, to NRC request for additional information dated November 19, 1992, you stated in Item No. 6, that Tennessee Valley Authority did not intend to change the licensing basis for BFN, Units 1, 2, and 3, prior to.
the receipt of the NRC staff's plant-specific safety evaluation.
However, in your
. Amendment No. 10, dated July 22, 1993, to the BFN Final Safety Analysis Report
,(FSAR), you have revised your licensing basis to use the guidelines and criteria of USI AC6 and associated Seismic Experience Database as an alternative method of equipment seismic qualification (Paragraph C.6.3.3 ofAppendix C to Updated FSAR dated July 22, 1993). As of the date of this letter, the staff has not issued its final SE on the USI A-46 implementation at BFN. The staff is unaware whether the GIP-2 procedure has been employed by TVAin actually making a change to the facilityoutside the scope of USI A46.
In order for the staff to complete its review of USI A-46 implementation at BFN, the staff r'equests that you (1) submit, for staff review, the complete documentation associated with your 10 CFR 50.59 evaluation for carrying out the FSAR changes for seismic qualification of equipment at BFN, and (2) identify any actual change to the system or components outside the scope of USI A-46, or any replacements and new equipment items for the facility using the approach described in Paragraph C.6.3.3 ofAppendix C to the Updated FSAR dated July 22, 1993.
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<s EEEEtlQW:
Letter, TVAto NRC, "Browns Ferry Nuclear Plant - Supplement 1 to Generic Letter 87-2, Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue A46 and Supplement 4 to Generic Letter 88-20, Individual Plant Examination of External Events for Service Accident Vulnerabilities," dated September 21, 1992.
2.
Letter, TVA to NRC, "Browns Ferry Nuclear Plant - Units 2 8 3 - Generic Letter 87-02, Supplement 1, Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue A-46 and Generic Letter 88-20, Supplement 4, individual Plant Examination of External Events for Service Accident Vulnerabilities -'ubmittal of Seismic Evaluation Reports," dated June 28, 1996.
3.
Letter, TVAto NRC, "Browns Ferry Nuclear Plant - Units 2 8 3-Generic Letter 87-02, Supplement 1, 120-day Response, Request for Additional Information," dated January 19, 1993.
II I'