ML20036C221
| ML20036C221 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/10/1993 |
| From: | Labarge D Office of Nuclear Reactor Regulation |
| To: | Medford M Tennessee Valley Authority |
| References | |
| NUDOCS 9306150311 | |
| Download: ML20036C221 (4) | |
Text
Docket Nos.
50-327 and 50-328 June 10, 1993 Tennessee Valley Authority ATTN: Dr. Mark 0. Medford, Vice President Technical Support 3B Lookout Place 1101 Market Street i
Chattanooga, Tennessee 37402-2801 l
Dear Dr. Medford:
l
SUBJECT:
REQUEST FOR TECHNICAL REVIEW OF DRAFT INFORMATION NOTICE REGARDING THE MAIN STREAMLINE BREAK ANALYSIS FOR MAIN STEAM VALVE VAULTS AT THE SEQUOYAH AND WATTS BAR NUCLEAR PLANTS
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As discussed with your staff, this letter forwards for your review a draft information notice regarding the main steamline break analysis for the main i
steam valve vaults at the Sequoyah and Watts Bar Nuclear Plants to verify the correctness of technical information.
Please return any comments you may have to me as soon as possible.
t Your cooperation is appreciated.
If no comments are received by June 25, l
1993, we will assume the technical information in the notice is correct.
i If you have any questions regarding this issue, please phone me at l
(301) SO4-1472.
Sincerely, Ociainal signed by, Drvid E. LaBarge, Senior Proj,ect Manager Project Directorate II-4 Division of Reactor Projects 0: fice of Nuclear Reactor Regulation
Enclosure:
Draft Information Notice cc w/ enclosure:
See next page Distribution Docket File NRC & Local PDRs SQN Reading SVarga GLainas FHebdon BCl ayton DLaBarge WTLeFave PTam BGrimes DORS R/F OGCB R/F GHMarcus 2
k' NECampbell CMcCracken OGC ACRS (10)
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Tennessee Valley Authority Sequoyah Nuclear Plant ATTN: Dr. Mark 0. Medford cc:
Mr. W. H. Kennoy, Director County Judge Tennessee Valley Authority Hamilton County Courthouse ET 12A Chattanooga, Tennessee 37402 400 West Summit Hill Drive Knoxville, Tennessee 37902 Regional Administrator U.S.N.R.C. Region II l
Mr. R. M. Eytchison, Vice President 101 Marietta Street, N.W.
Nuclear Operations Suite 2900 Tennessee Valley Authority Atlanta, Georgia 30323 3B Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801 Mr. William E. Holland Senior Resident Inspector Sequoyah Nuclear Plant Mr. M. J. Burzynski, Manager U.S.N.R.C.
Nuclear Licensing and Regulatory Affairs 2600 Igou Ferry Road Tennessee Valley Authority Soddy Daisy, Tennessee 37379 5B Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801 Mr. Jack Wilson, Vice President Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, Tennessee 37379 TVA Representative Tennessee Valley Authority 11921 Rockville Pike Suite 402 Rockville, Maryland 20852 Ms. Marci Cooper, Site Licensing Manager Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 i
Soddy Daisy, Tennessee 37379 Mr. Michael H. Mobley, Director Division of Radiological Health t'
3rd Floor, L and C Annex 401-Church Street Nashville, Tennessee 37243-1532 General Counsel Tennessee Valley Authority i
ET llH 400 West Summit Hill Drive Knoxville, Tennessee 37902
4-i UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.
20555 June
, 1993 NRC INFORMATION NOTICE 93-XX:
POTENTIAL PROBLEM WITH MAIN STEAMLINE BREAK ANALYSIS FOR MAIN STEAM VAULTS / TUNNELS Addressees All holders of operating licenses or construction permits for pressurized water reactors.
Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this-information -
notice to alert addressees to a potential inadequacy in the main steamline break analysis which could place some pressurized-water reactor (PWR) plants outside their current structural design basis for the main steam valve vaults or main steam tunnels. The plants of concern are those that must postulate a double-ended rupture of a main steamline in these areas'.
It is expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required.
Description of Circumstances i
During the Watts Bar Calculation Reconstitution Program, Tennessee Valley Authority (TVA) discovered that Westinghouse had supplied nonconservative data
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for the main steamline break analysis which could result in the structural design margins being exceeded in the main steam valve vaults. TVA had requested Westinghouse to reevaluate the 1975 Westinghouse mass and energy -
release data used in the Watts Bar analysis for these valve vaults, and to advise TVA if the data were still applicable. _0n June 23,1992, Westinghouse advised TVA that the 1975 mass and energy release data were no longer considered conservative, and were not applicable for a pressure transient evaluation of the vented main steam valve vaults.
Failure to account for liquid entrainment in the blowdown, resulted in a reduced mass and energy release rate in the 1975 data.
Westinghouse then provided a bounding analysis based on ANSI /ANS Standard 58.2 (1980) methodology which included liquid entrainment in the blowdown. This new analysis indicated that the valve vault structural design pressure would i
be exceeded.
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DRAPT IN 93-xx June xx, 1993 Page 2 of 3 On August 17, 1992, for Sequoyah Nuclear Plant (1.ER 50-327/92-013), and on October 13, 1992, for Watts Bar Nuclear Plant (CDR 50-390/92-09), TVA reported the use of nonconservative Westinghouse data for the main steamline break analysis which could result in the valve vault structural design pressure being exceeded. TVA had determined that the mass and energy release data for Watts Bar were also applicable to Sequoyah.
TVA prepared a justification for continued operation (JCO) for Sequoyah.
This JC0 will be in effect until the startup from refueling Cycle 6 for both Sequoyah Unit 1 (Fall 1993) and Unit 2 i
(early 1994).
DISCUSSION The 1975 mass and energy release data. supplied by Westinghouse was based upon the largest steam generator depressurization rate consistent with a high-quality steam discharge. The Westinghouse data was applicable for a postulated double-ended main steamline break in the turbine building, assuming flow in both the forward and reverse direction. Apparently, TVA applied this data without verifying its applicability to vented compartments, such as the main steam valve vaults. A dry steam release in a vented compartment such as the main steam valve vault may not be conservative, because of moisture entrainment within the discharge.
For a main steamline break analysis, the limiting plant conditions for the steam generator mass inventory and secondary system pressure are often at hot standby / shutdown plant conditions (0 percent power level, primary plant at operating temperature and pressure). Due to the high flow rates associated with the main steamline break, frothing in the steam generator raises D water level rapidly, which decreases the quality of fluid expelled frot the steam generator. Although the enthalpy of this lcw-quality fluid is less than the enthalpy of dry steam, the critical mass flow is 4 to 5 times higher, resulting in a net increase in the energy release rate from the break.
This may be the limiting case for determining maximum pressure in vented (blowout panels) compartments.
Westinghouse recommended to TVA that the methodology outlined in ANSI /ANS Standard 58.2 (1980), Appendix E be used to generate the Watts Bar bounding mass and energy release rates,' which would determine the pressure inside the valve vaults. This mass and energy release data would include the entrainment of water and bound the analyses that could be conducted for this type of event. The NRC staff agrees with the Westinghouse statements about the ANSI /ANS 58.2 methodology. Westinghouse performed the Watts Bar analysis with the ANSI /ANS 58.2 methodology and informed TVA of a significant increase in the mass and energy releases generated over those of the original analysis.
j TVA determined that the increased Watts Bar mass and energy release rates j
produced pressures that exceeded the present structural design margins, and i
challenged the structural adequacy of the walls and. slabs of the main steam valve vaults. TVA calculations showed that the peak pressures in the valve vaults could increase by about one-third when moisture entrainment was i
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&y IN 93-xx June xx, 1993 Page 3 of 3 considered.
Failure of the valve vault walls or slabs could damage such equipment as main steam system, main feedwater system, and auxiliary feedwater system components and piping. This equipment damage could result in the inability (or reduced ability) to feed the intact steam generators, or in the blowdown of more than one steam generator.
Upon consultation with Westinghouse, TVA determined that the analysis data for the Sequoyah main steam valve vault rooms were also nonconservative. A JC0 has been prepared for Sequoyah.
The JC0 is based on the Sequoyah main steam system piping design in the valve vaults meeting most of the break exclusion provisions of the Standard Review Plan (SRP) Branch Technical Position (BTP)
HEB 3-1, " Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment." A postulated one-square-foot break was analyzed for the JC0 interim period. The revised calculated pressures (using the ANSI /ANS 58.2 methodology) were bounded by the original design pressure of the vaults. This JC0 will be in effect until the next Sequoyah, Units I and 2 refueling outages (Cycle 6 for both units).
TVA will make plant modifications to bring the plant into compliance with the original design basis.
The modifications will involve modifying each of the fluid head anchor-sleeve openings to decrease the flow area in the event of a postulated break, thereby limiting the mass and energy release rate.
The flow area will be sized to limit the pressure in the main steam valve vaults to less than the original design basis of the floor and walls.
Combustion Engineering and Babcock & Wilcox designed PWRs may also be affected by this issue if vected compartments have been analyzed nonconservatively, assuming dry ste4m. Therefore, this information notice is being sent to all PWR licensees r,d holjers of PWR construction permits.
This informatiu not4_i: rquires no specific action or written response.
If you have any quutions about the information in this notice, please contact one of the techniel contacts listed below or the appropriate Office of Nuclear Reactor Reguistion project manager.
Brian K. Grimes, Director Division of Operating Reactor Support Office of Nuclear Reactor Regulation Technical contacts:
J. B. Brady, Region II (404) 331-0339 W. T. Lefave, NRR (301) 504-3285
Attachment:
List of Recently Issued NRC Information Notices