ML20214N620

Response to FOIA Request.App E Documents Re NRC Insp Rept 50-443/84-07 Already on File in PDR

ML20214N620
Person / Time
Site:	Seabrook
Issue date:	11/20/1986
From:	Grimsley D NRC OFFICE OF ADMINISTRATION (ADM)
To:	Curran D HARMON & WEISS
References
FOIA-86-756 NUDOCS 8612030701
Download: ML20214N620 (2)

See also: IR 05000443/1984007

Text

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U.S. NUCLEAR REXULATORY COMMISSION Nac FOia u utST N ustaSi

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I INFORMATION ACT (FOIA) REQUEST

N eeee: NOV 2 0198C

DOCKET NUMsER.Si arf espwaamp

REQUESTER , ,

$hh h & Ib

PART f.-RECOR / RELEASED OR NOT LOCATED (See checked bones!

No agency records subrect to the request have been located.

No additonal agency recorda subpect to the request have been located.

Agency records subrect to the request that are identifed in Appendix

1717 H Street, N.W., Washington, DC.

h are already avoitable for public inspecten and copying in the NRC Public Document Room,

Agency records subject to the request that are identifed in Appendia are being made available for public inspecten and copying in the NRC Public Document

Room,1717 H Street, N W., Washington, DC, in a folder under this FOIA number and requester name.

The nonproprietary version of the proposalist that you agreed to accept in a telephone conversaton with a member of my staff is now being made avalable for publec inspection

and coving at the NRC Public Document Room,1717 H Street, N W , Washington, DC, in a folder under this FOIA number and requester name.

Enclosed is informaton on how you may obtain access to and the charges for copying records piaced in the NRC Public Document Room,1717 H Street, N.W., Washington, DC.

Agency records subrect to the request are enclosed. Any applicable charge for copies of the records provided and payment procedures are noted in the comments secton. l

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Cocords subl ect to the request have been referred to another Federal agencytes) for revew and direct response to you. l

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In view of NRC's response to the request. no further acton is being taken on appeal letter dated

PART ll.A-INFORMATION WITHHELD FROM PUBLIC DISCLOSURE

Certain informaton in the requested records is being withheld from public disclosure pursuant to the FOIA enemptions desenbod in and for the reasons stated in Part it, sec-

tsoro B, C, and D. Any released portons of the documents for which only part of the record is being withheld are being made avadable for public inspection and copying in

the NRC Public Docurr.ent Room,1717 H Street, N W., Washington, DC, in a folder under the FOIA number and requester name.

Comments

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8612030701 861120

PDR FOIA

CURRAN 86-756 PDR

SIG - *

E. DIRECTOR. Olvisi 5 RULES RECORDS

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APPENDIX E

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RECORDS MAINTAINED AMONG POR FILES

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NUMBER DATE DESCRIPTION

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1. 06/28/84 NRC Inspection Report 50-443/84-07 - ANO 8408170117

(101 pages)

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. HARMON & WEISS

2000 $ STREET N.W.

SutTE 430

WASHINGTON, D.C. soooo-nes

GAIL McOREEVY HARMON

ELLYN R.WEIS$ TELEPHONE

OlANE CURRAN (702)328-3500

DEAN R. YOUSLEY

ANDREA C. FERSTER

October 21, 1986

BAND DELIVERED

FREEDOM OF INFORMATION

Donnie H. Grimsley, Director ACT REQUEST

Division of Rules and Records '[b" 75[O

of fice of Administration

U.S. Nuclear Regulatory Commission

7735 Old Georgetown Road & #d / ~2/~/6

Be thesda, Ma ryland

Dear Mr. Grimsley:

Pursuant

et seq., to the Freedom of Information Act, 5 U.S.C. S 552,

the New England Coalition on Nuclear Pollution ("NECNP")

requests that you make available all documents that relate to the

cadweld test and/or construction problem referred to in paragraph

8 of a letter from Stephen M. Long, NRC, to Robert J. Barrison,

Public Service Company of New Hampshire, dated October 8,1986.

A copy of that letter is attached. Please note that this request

may overlap paragraph 4 of another FOIA

September request made by NECNP on

12,1986 (FOIA 8 6-678) .

proceeding. NECNP is an intervenor in the Seabrook operating license

The organization intends to use this information in

the licensing hearings to further the public's interest in the

safety of operation of the Seabrook plant. Therefore we request

that you waive any copying and search fees pursuant to 10 C.F.R.

9.14(c).

I

days, look forward to receiving your response within ten working

as required by the Freedom of Information Act.

Since rely ,

.

! Diane Curran

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RECElVED OCT 17 f986

UNITED STATES

!N , , 'g NUCLEAR REGULATORY COMMISSION

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g OCT $66

Docket P:os.: 50-443 *

and 50-444

Fr. Robert J. Harrison

President & Chief Executive Officer ~ ~

Public Service Company of hew Hampshire

Post Office Box 330

Manchester, New Hampshire 03105

Dear Vr. Harrison:

Subject: Fequest for Additional Information for Seabrook Station, Units 1 and

2, Errergency Planr.ing Sensitivity Study

The enclosed Request for Additional Information dccuments the oral and handwritten

questions transmitted to Public Service Company of New Hampshire personnel and

contracturs during our meeting in Bethesda, Maryland on September 23, 1986.

Please provide your responses promptly to facilitate our review.

Questions or additional information regarding this matter should be directed to

the Techr ical Project Manager for the review of the Seabrook Emergency

Planning Sensitivity Study, S. M. Long (301) 492-8413.

Sincerely,

) D5

Steven M. Long, Project Manager

PWR Project Directorate No. 5

Division of PWR Licensing-A

Enclosure:

As stated

cc: See next page

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,y Enclosure

REQUEST FOR ADDITIONAL INFORMATION

SEABROOK STATION, UNITS 1 AND 2

DOCKET h05.: E0-443 AhD 50-444

EMERGENCY PLANNING SENSITIVITY STUDY

1. Describe how the overpressurization calculations made by SMA were checked

or design reviewed.

2. A meeting should be arranged with the originator of these calculations

to assist the BNL reviewers in following these calculations and understanding

the assumptions.

3. Eccument the basis for the assumptions in the calculations. In particular,

explain the uncertainty factors assigned to various pressure capacities.

4 Explain the mechanism for transferring the load from the penetration

sleeves to the containment wall, in particular, the equipment hatch, when

subjected to high strain conditions. Explain how the rebars around the

penetrations were assessed to assure that they can resist these loads in

addition to the primary pressure induced loads.

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5. The calculations use a rebar ultimate stain value of 4.7%, i.e., more than

21 feet of linear extension for the hoop bars. This linear extension

under the high pressure load will be accommodated by formation of cracks in

the concrete totaling approximately 21 feet in width. Justify the assumption

that the pressure loads will be carried proportionately by the linear plate

and the rebars (similar to the elastic condition) in this highly cracked

condition. Also address the potential for developing a crack large enough

for the local extension of the liner plate to lead to its failure at that

point.

6. Was compatibility of strains in the rebars and the liner plate satisfied

in the calculations? For example, the outermost hoop bars will fail before

the inside bars and the liner plate reach their respective ultimate strengths.

Was this fact reflected in the calculations? In addition, how is the

biaxial stress-strain state of the liner plate considered.

7. The combined tension, shear and bending effect at base and spring line

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levelswasnotconsideredinthecalculations(Ref.p.35, assumption 6).

i Verify'that the combined effect does not change the conclusions of the

l analysis.

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. Since 31 cadwelds out of a total of 169 test samples failed at a stress

lower than the rebar ultimate strength and there was apparently a i

construction problem concerning staggering of these welds. provide "

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\~~. justification for not usinjLa__ reduced-ulttma'te strengt6 for, the rWar.

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9. The containment analysis is based on an axisynnetric geometry and loading.

This is not the case due to the presence of adjoining structures such as

the fuel building and main steam and feedwater pipe chase. Identify these

axisynnetric conditions and assess their impact on the failure modes and

analysis.

10. Only a sample of pipe penetrations are considered in some detail (X-23,

X-26andX-71). The justification to consider only these should be

provided.

11. A structural evaluation of electrical penetrations should be provided.

12. The basis for the leakage area assigned to the flued head at failure

should be provided.

13. A more detailed evaluation of the impact of punching shear at the Fuel

Transfer Building should be provided.

14 Clarify the extent to which double ended piping failures have been

considered in the overall containment performance assessment. Provide

isometric drawings of all piping attached to containment penetrations.

15. In PLG-0465, page 2-10, Figure 2-3, the conditional frequency of

exceeding whole body dose vs distance appears to be driven by the S2

source term. If this is the case, please describe all accident sequences

(internal and external events) that contribute to the frequency of the

52 source term given in Table 4-2, pg. 4-7. in particular define how ,

the timing and size of containment leakage was determined for each of

these classes of accident sequences. Justify the appropriateness of the

binning of each of the accidents into this particular source term.

16. Provide justification for the liner yield stress increase from the

specified yield stress of 32 ksi to a mean yield stress of 45.4 ksi.

17. Indicate the correlation between containment failure sequences and the

l containment failure modes.

18. Provide the basis for concluding that the sight glasses in the hatches

! will not fail under high containment temperature and pressure conditions.

l 19. Document the effect that the recent update in seismic fragilities will

have on the conclusions of the PSA results.

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. 20. Assess the impact on risk of using the assumption of ultimate containment

capability predicted by VE&C analysis (150 psig). E

21. What is the impact on risk from accidents during shutdown and refueling

when the containment function may not,be available?

2E. It is the staffs understanding that preexisting violations of containment

integrity were " included" in the PSA by assuming the average effect was

to raise the containment leak rate to the design basis value of 0.15/ day.

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a. Compare this assumption with the containment integrity violation

data presenttd in huREG/CR 4220.

b. Khat contributions would these containment integrity violation data

make to the probabilities for each of the release categories (Assume -

the 55W category is redistributed over all the appropriate categories

, by the conditional probabilities of preexisting leakage paths of the

size appropriate to each category).

23. a. Provide a narrative description that quantitatively delineates the

dominant contributors to the dose probability vs distance curves and

the early fatality probability curves. The dominant release

categories should be specified and the dominant accident sequences

. contributing to each of these release categories should be specified.

The probability of occurrence of each release category should be

stated. These data should be provided for the current study and

for the original PSA results. Changes between the two studies i

should be attributed to specific differences in the analysis.

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b. Provice a set of early fatality conditional probability curves for

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each release category, assuming evacuation distances of 1 mile and

2 miles.

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c. Provide the conditional mean risk of early fatality for each of 1

the curves provided in b. '

l 24. Provide a quantitative description of the effects of the following differences i

i between the original PSA and the current study:

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l a. reduction in probability of core-melt V sequences

l b. factor of 1000 scrubbing of releases through RHR seals

i c. change of release category (56 to SI) for unscrubbed event V sequences.

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l The effects should be described in terms of differences in risk curves

for early fatalities and fur 200 rem vs distance. .

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25. Provide a list of all paths for loss of RCS inventory outside containment. I

Show how these have been considered with respect to LOCA and with respect

to containment bypass for radioactive materials following core damage.

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26. Indicate the extent to which the effect of local deflagration / detonation

of hydrogen qas concentration in localized areas both inside and outside

the containment has been considered in the assessment of risk. Include

a discussion of how weak areas of containment have been considered in

your assessment, for example, the containment is considerably weaker

in its resistance to pressure loading from outside the containment.

27. Discuss the effect on risk of hydrogen deflagation/ detonation in the RHR

vault.

22. Identify any penetrations connected directly into the containment atmosphere

which rely on any remote manual or manual valves for isolation.

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