ML20207A540
| ML20207A540 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/29/1999 |
| From: | Colburn T NRC (Affiliation Not Assigned) |
| To: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| References | |
| GL-96-06, GL-96-6, TAC-M96877, NUDOCS 9905270014 | |
| Download: ML20207A540 (4) | |
Text
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a UNITED STATES g
NUCLEAR REGULATORY COMMISSION
't WASHINGTON, D.C. 20665-0001
,o April 29, 1999 s
Mr. James W. Langenbach, Vice President and Director, TMl GPU Nuclear, Inc.
P.O. Box 480 Middletown, PA 17057
SUBJECT:
GENERIC LETTER 96-06 RESPONSE FOR THE THREE MILE ISLAND NUCLEAR STATION, UNIT 1 (TMI 1) (TAC M96877)
Dear Mr. Langenbach-Generic Letter (GL) 96 06, " Assurance of Equipment Operability and Containment Integrity i
During Design-Basis Accident Conditions," dated September 30,1996, included a request for licensees to evaluate cooling water systems that serve containment air coolers to ensure that they are not vulnerable to waterhammer and two-phase flow conditions. Additionally, licensees were requested to evaluate piping systems that penetrate containment to determine if they are susceptible to thermal expansion of fluid which could lead to overpressurization of piping.
You provided an assessment of the waterhammer, two-phase flow, and thermally-induced piping overpressurization issues for TMI-1 and responses to the staff's requests for additional i
information dated February 14 and June 2,1997, and February 13,1998. A response to the staff's July 13,1998, request for additional information was provided in a letter dated September 30,1998. Based on the staff's review of the information in your submittals, it is our
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understanding that GPU Nuclear, Inc., has performed bounding analyses that are consistent with the design and licensing bases for TMI 1 in order to address the waterhammer and two-phase flow concems. Your analyses indicate that the worst-case scenario would result in a void faction of about 0.12 percent within the fan coolers, which is so small that any resultant waterhammer or two phase flow effects would be inconsequential. In conjunction with the analyses that were performed, you also made the following changes to the reactor building emergency cooling (RBEC) system:
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. the RBEC system relief valves were moved outside containment to address ficoding concems, and f
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the fan cooler discharge back pressure control valves were connected to the 2-hour, safety-a grade, backup air system to assure that these valves will be able to function during a loss of instrument air.
The staff has determined that the analytical methodology, assumptions, and system modifications as described in your September 30,1998, submittal appear to be reasonable and appropriate for the specific design and configuration of the RBEC system at TMI-1, and the staff is satisfied with GPU Nuclear, Inc.'s resolution of the waterhammer and two-phase flow elements of GL 96-06 for TMI-1.
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1 J.Langenbach -
4 April 29, 1999 e
1 Wdh respect to the GL 96-06 issue conceming thermal overpressurization, in your submittal of February 14,1997, you identified 11 penetrations potentially vulnerable to a water solid volume that may be subjected to an increase in pressure due to heating of the trapped fluid.
The affected 11 lines are: "A" and B" Once through Steam Generator Sampling Lines; intermediate Closed Cooling Water Retum Line; Reclaimed Water Supply Line; Makeup and Purification Letdown Outlet Line; Pressurizer and Reactor Coolant Sampling Line; Reactor Coolant Pump Seal Water Retum Line; Reactor Coolant Drain Tank Transfer Line; Reactor Coolant Pump Cooling Retum Line; and "A" and "B" Core Flood Tank Sampling Lines. You' determined that the affected lines are operable based on the criteria in Appendix F of Section lli of the ASME Code. For your long-term corrective action, you committed to install pressure relief devices on the affected lines. Your June 2,1997, submittal stated that based on its further analyses, stresses in the Reactor Coolant Pump Seal Water Return Line met the design basis allowable stresses.
I la your February 13,1998, submittal you stated that you had installed pressure relief devices on 10 affected piping segments during the September 1997 refueling outage. In your September 30,1998, response to the staff's request for additional information, you provided design basis information for the Reactor Coolant Pump Seal Water Retum Line. The staff finds the design basis for the Reactor Coolant Pump Seal Water Return Line to be reasonable and acceptable.
The staff finds that GPU Nuclear, Inc.'s corrective actions provide an acceptable resolution for the issue of thermally-induced pressurization of piping runs penetrating the containment.
This completes the staff's review of GL 96-06 for TAC No. M96877. If you have any questions, please call me at (301) 415-1402.
Sincerely, ORIGINAL SIGNED BY:
Timothy G. Colburn, Senior Project Manager, Section 2 Project Directorate i Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-289 cc: See next page plSTRIBUTION Docket File i OGC PUBLIC ACRS i
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-J.Langenbach With respect to the GL 96-06 issue concerning thermal overpressurization, in your submittal of February 14,1997, you identified 11 penetrations potentially vulnerable to a water sohd volume that may be subjected to an increase in pressure due to heating of the trapped fluid.
The affected 11 lines are: "A" and "B" Once through Steam Generator Sampling Lines; intermediate Closed Cooling Water Return Line: Reclaimed Water Supply Line; Makeup and Purification Letdown Outlet Line; Pressurizer and Reactor Coolant Sampling Line; Reactor j
Coolant Pump Seal Water Return Line; Reactor Coolant Drain Tank Transfer Line; Reactor j
Coolant Pump Cooling Return Line; and "A" and "B" Core Flood Tank Sampling Lines. You determined that the affected lines are operable based on the criteria in Appendix F of Section ill of the ASME Code. For your long-term corrective action, you committed to install pressure relief devices on the affected lines. Your June 2,1997,' submittal stated that based on its further analyses, stresses in the Reactor Coolant Pump Seal Water Return Line met the design basis allowable stresses.
In your February 13,1998, submittal you stated that you had installed pressure relief devices on 10 affected piping segments during the September 1997 refueling outage. In your September 30,1998, response to the staff's request for additional information, you provided design basis information for the Reactor Coolant Pump Seal Water Return Line. The staff finds the design basis for the Reactor Coolant Pump Seal Water Return Line to be reasonable and acceptable.
The staff finds that GPU Nuclear, Inc.'s corrective actions provide an acceptable resolution for
- the issue of thermally-induced pressurization of piping runs penetrating the containment.
This completes the staff's review of GL 96-06 for TAC No, M96877. If you have any questions, please call me at (301) 415-1402.
Sincerely, Me i
Timothy G. C51 burn, Senior Project Manager, Section 2 Project Directorate l Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-289 l
cc: See next page l
Three Mile Island Nuclear Station, Unit No.1 i
cc:
Michael Ross Director, O&M, TMl Robert B. Borsum GPU Nuclear, Inc.
B&W Nuclear Technologies P.O. Box 480 Suite 525 Middletown, PA 17057 1700 Rockville Pike Rockville, MD 20852 John C. Fomicola Director, Planning and William Domsife, Acting Director Regulatory Affairs Bureau of Radiation Protection GPU Nuclear, Inc.
Pennsylvania Department of i
100 Interpace Parkway Environmental Resources I
Parsippany, NJ 07054 P.O. Box 2063 Harrisburg, PA 17120 Jack S. Wetmore Manager, TMi Regulatory Affairs Dr. Judith Johnsrud GPU Nuclear, Inc.
National Energy Committee i
P.O. Box 480 Sierra Club Middletown, PA 17057 433 Orlando Avenue State College, PA 16803 Ernest L. Blake, Jr., Esquire Shaw, Pittman, Potts & Trowbridge Peter W. Eselgroth, Region 1 2300 N Street, NW.
U.S. Nuclear Regulatory Commission Washington, DC 20037 475 Allendale Road King of Prussia, PA 19406 Chairman Board of County Commissioners of Dauphin County Dauphin County Courthouse Harrisburg, PA 17120 Chairman 1
Board of Supervisors of Londonderry Township R.D. #1, Geyers Church Road Middletown, PA 17057 Wayne L. Schmidt Senior Resident inspector (TMI-1)
U.S. Nuclear Regulatory Commission P.O. Box 219 Middletown, PA 17057 Regional Administrator Region i U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406