ML20212H499

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Forwards Summary of Major Events Occurring on Day 31 from B&W Trial Records.Questions Re Steam Generators Raised
ML20212H499
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 09/30/1986
From: Davenport D
AFFILIATION NOT ASSIGNED
To: Blough A
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20212H217 List:
References
NUDOCS 8701210461
Download: ML20212H499 (5)
v • d • e


Text

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[TO:;

Allen R. Blough, #U P E 'I N'6 i

Reactor 1-rojects, Section 1 A, ,

! Division of Reactor Projects, 1:RC Rocion 1,

.-631 Park Avenue,liing of Prussia, Pc.

Dear I.r. Blough:

l' As I taentioned to you in the beginninC of the Unit 1 Salp Meeting, l'm forewarding the identifying data on the Day 31 I page fror. the B&W Trial records. The.page in questi~4 is 1

the H. L. Long Exibits. . .Exibit I! unbar 692. According . to the i l

l .

I cc s in the accompanying cover letter, a copy of the exibit, l a sequence of events, and the letter were sent to a E.H. Grier of Region l's Office of Inspection and Enforcen:ent, so you may '

L have a full copy sortetthere in Region 1..it also went to an S. l. .

Varga in D.C..

I'nclosed is a xorox of tne cover letter.

The Day 31 ovents did rt cc some additional questions regm ding q steam generators whcu primary to secondary contamination of note  !

occurs. . . .who authorizos the continued use. . .or putting bach in use of said generators..and what arc. the fac tors governing such decisions. If I understood correctly...there has boon som initial study or decision underway which advocates the steauing on the dataged generator, rather than its isolation...and that by so doing... releases would puroortedly be lower to the envirumont outside the plant. Could you comuont on 'that again, or could . c.

Conte..I believe he mentioned the study in progress. I uay nt>t havo urAerstood the comuerts made're. the proceeddres to be followed regarding ci 3nificant steam tube leahs, e' steau tube failuren.

Of course, it uust be assumed that all tube failures riould be s : ;nific.,nt.

8701210461 870113

.PDR ADOCK 05000289.

5 PDR

l .

(2)

- Alco note on day 32, the B Generator was apparently icainted CCri">

because i.leasurenento on the station vont uonitor increased.

Actuall'y....the Turbino Bypass valvoc .:s9: isolated. Would this cean B was isolated totally re. rol cosec? The readinco on the 32nd day were felt to have -been due :to the changing of a charcoal filter; however, one might at1.11 wonder about the cauce of' the increased levels of radioacts ve materials present in Ti:I 2 on the 31st day of the accident. , 'Am also wondering what stack monitor read in the 31st day, also ir sooo defect eninto, or potential for added releases is present 17 condensore thet would have caused releases like those noted on Day 317.

t since rely, d/%e loborah Davenport, 1002 i.artet Stree t ,

Casp I:lll, Penna. 1701i 1- 717- 763-952 Also note, are copies of my letters, end the nerces en vell 7 being sent to the 1D2c and the service lictc, ; onca not been able to Get to document room, and was not certain if all data rac boing cent through. If it has not included, I would like to requent that all letters and xeroxes be sont out to all of lists etc...

if co I thank you retrospectively.

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$31 PARK AVENUE KING OF PRUS$1A, PENNSYLVANI A 19406

~;

/

JAN 1 b 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, LI 'N.f** '

Allen R. B16 ugh, Chief Projects Branch No. 1 Division of Reactor Projects Attachments:

As Stated cc: .

PDR LPDR Senior Resident Inspector (TMI-1) bec:

A. Blough R. Conte W. Baunack K. Abraham pat 9lO 30T 4@'

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 (LPOR). Specifically Sections 4.3 of both documents address this topic along with licensee procedures for handling 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

. ~

[4 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 (from the normal 579* F); and an orderly cooldown and depressurization is commenced. The affected steam generator is isolated if the off-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-2?

Response

In response to your specific question about the turbine bypass valve, a number of valves in the main steam and feedwater system must be shut to 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-I 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.

9 1

.; o t

i.

3

o

.P

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 POR and LPDR, 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.

1

.

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