NUREG-1019, Forwards Response to 860915 & 30 Questions Re Steam Generator Tube Rupture & Turbine Bypass Valve
| ML20212H211 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/13/1987 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Davenport D AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML20212H217 | List: |
| References | |
| RTR-NUREG-1019, TASK-1.C.1, TASK-TM NUDOCS 8701210305 | |
| Download: ML20212H211 (4) | |
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JAN 13.1987 Docket No. 50-289 Ms. Deborah Davenport 1802 Market Street Camp Hill, Pennsylvania 17011
Dear Ms. Davenport:
This is in response to your letters, dated September 15 and 30,1986, to me. We have summarized our responses to questions in your letters in the attachment. Your letters are attached for completeness and forwarding to the docket rooms.
If you have questions, please contact either Rich Conte (215-337-5146) or me.
Sincerely, Original SIEned E g f
' Allen R. Blough, Chief Projects Branch No. 1 Division of Reactor Projects Attachments:
As Stated cc:
PDR LPDR Senior Resident Inspector (TMI-1) bec:
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ATTACHMENT RESPONSES TO D. DAVENPORT LETTERS A.
Letter dated September 15. 1986 1.
In reference to releases during a steam tube failure, provide copies of any evaluation or procedures that are available.
Response
The design basis steam generator tube rupture event was evaluated in relation to the Kinetic Expansion repair process for the TMI-1 steam generator tubes. The results of that review are documented in NUREG 1019 (Section 4) and Supplement 1 (Section 4), which should be avail-able for review in the Local Public Document Room (LPDR).
Specifically Sections 4.3 of both documents address this tcpic along with licensee procedures for h'andling such events.
Further, the licensee's method-ology for handling such events are being generically reviewed under TMI Task Action Plan (TAP) Item No. I.C.1, " Accident and Procedure Review." Recent licensee submittals on this topic should also be available for review in the LPDR.
2.
Several sections of the letter reiterate the request for a filter on the condenser off gas system.
Response
This request was answered in a letter, dated November 26, 1986, from H. Denton, NRR, to you.
B.
Letter dated September 30, 1986 1.
Comment on the steaming of, rather than isolating, a damaged steam generator (upon tube rupture) and how that lowers releases to the environment.
Response
If a steam generator tube rupture should occur along with multiple tube ruptures, the best way to stop the ensuing primary to secondary leakrate is to cooldown and depressurize the reactor coolant system The quickest way to do that within cooldown limits is to use both steam generators rather than other slower methods such as steaming one undamaged steam generator or using the RCS feed and bleed meth-odology. Use of the steam generator is complicated by releases of radioactivity eventually through the condenser off gas system or through the steam generator safety / relief valves if the leak is large enough to cause RCS pressure (if above 1040 psig) to be applied to the secondary side of the steam generator (relief valve settings at 1040-1100 psig).
Releases are inevitable but they are within 0FFICIAL RECORD COPY 0223 - 0002.0.0 01/12/87
2 design basis limits (10 CFR 100) for the design basis event (single tube failure). The licensee is also prepared for the beyond-design-basis event (multiple tube failure) in order to minimize releases to the public.
For any of the events noted above and those releases, there will be a dose rate to the public. The licensee's approach is to shorten, as much as possible, the length of time to stop the pri-mary to secondary leakage, so that the total dose to the public will be lessened.
This approach is under generic review by NRC as stated in paragraph A.1 above.
Briefly, upon detection of primary to secondary leak rate of 1 to 50 gpm, licensee procedures require the plant to be shutdown and immedi-ately cooled to below 540 F (frem the normal 579 F); and an orderly cooldown-and depressurization t commenced. The affected steam generator is isolated if the oft-site dose projection approaches 50 mR/hr whole body or 250 mR/hr thyroid dose rate.
2.
In reference to day 32 after the TMI-2 accident, Ms. Davenport asks that, if the turbine bypass was isolated, would that mean the "B"
steam generator would be isolated totally with respect to releases?
For day 31, what were the stack monitor releases at TMI-27
Response
In response to your specific question about the turbine bypass valve, a number of valves in the main steam and feedwater systen must be shut te isolate a steam generator; e.g., main steam isolation valves and feedwater isolation / block valves.
Even with this isolation, there is a possible release path from an isolated steam generator through the steam generator safety / relief valve, which could actuate auto-matically on a high pressure situation in the steam generator (at approximately 1040-1100 psig).
Even with the design, as noted in the NRR letter to you dated November 26, 1986, the TMI-1 design meets the objective of 10 CFR 50 Appendix I.
The TMI-2 accident and subsequent days after that event have been extensively reviewed and evaluated by NRC staff. It would not be an appropriate expenditure of NRC staff resources to revisit and answer specific questions about those events unless a new safety question needs to be addressed. Your request for a filtration system on the condenser off gas system, has been reviewed by NRC staff and answered by Mr. Denton's letter.
0FFICIAL RECORD COPY 0223 - 0002.1.0 01/12/87
3 3.
Are copies of my letters being sent to PDR's and service lists? If not, I am requesting that they be.sent to all lists, etc.
Response
All of,your letters to Region I are being sent to the PDR and LPOR, either separately or by attachment to the NRC staff response letters.
Any discrepancies in that regard should be brought to our attention.
With respect to the service list, it is not-the policy of Region I to serve this type of correspondence on the Board or parties.
OFFICIAL RECORD COPY 0223 - 0004.0.0 01/12/87