ML20214K190
| ML20214K190 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 08/11/1986 |
| From: | Speis T NRC |
| To: | Novak T NRC |
| Shared Package | |
| ML20214K191 | List: |
| References | |
| FOIA-86-678 NUDOCS 8608190580 | |
| Download: ML20214K190 (6) | |
Text
- ** " $t ENCLOSURE 2
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fg, UNITED STATES hsCLEAR REGULATORY COMMISSIOi, E.,
wassiNo TON. O. C. 20555
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AUG 11 !!B6 NOTE TO: Tom Novak FROM:
inemis Speis
SUBJECT:
REVIEW OF THE SEABROOK EPZ SUuni TAL I have reviewed your proposed memo to H. Denton on the Seabrook review and provided comments in the form of markups to V. Noonan on August 8.
It is important to decide what direction NRC is going to take on this issue before a detailed technical review can start. A decision chart set up in the fonn of three questions is attached for your consideration.
I would recommend that you assemble a small group to assess the potential approaches to the review.
Four individuals, one from each, PWRL, USRO, IE and OGC, could do the job in about twc seeks. Our representative is Len Soffer, please feel free to contact him directly.
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H. Denton j
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Potential Approaches to the Review of the Seabrook EPZ Submittal,
1.
Assuming that all technical information received from Seabrook is correct, can NRC reduce the Seabrook EPl or evacuation zone below the 10 mile limit under current regulations and established regulatory practice?
If answer is No, go to Q-2.
If answer is Yes:
What information is important for the decision?
What additional information is needed from PSNH?
2.
Can NRC use a risk. based criteria to justify a reduction in EPZ or evacuation zone for Seabrook without "rulemaking" or granting an exemption?
If answer is No - Seabrook should either join the ongoing rulemaking or go to y-3.
If answer is Yes:
What information is important for the decision?
What additional information is needed for PSNH7 3.
What basis could NRC have for granting an exemption from existing emergency planning requirements for Seabrook?
a.)
Is the Seabrook plant significantly different from other PWRs with large dry containments with respect to Emergency Planning requirements?
If answer is No, go to Q-3b.
If answer is Yes:
What information is important for the decision?
What additional information is needed from PSNHY b.)
Is there an "inmediate need" for Se' brook, that would justify a
exempting it from current EPZ requirements while the rule change is underway?
If answer is No - give up!
If answer is Yes:
i What information is important for the decision?
What additional information is needed from PSNH7
ENCLOSURE 3
References:
l 1.
George S. Thomas to Vincent S. Noonan letter dated July 29, 1986, Seabrook Station Probabilistic Safety Study Update.
2.
John DeVincentis to Vincent S. Noonan letter dated July 21, 1986, Seabrook Station Probabilistic Safety Assessment Update.
3.
Seismic Fragilities of Structures and Components of the Seabrook Generating Station, Units 1 & 2, prepared by NTS Engineering, Long Beach, CA for Pickard, Lowe and Garrick, Inc. and New Hampshire t
Yankee Division, Public Service Company of New Hampshire, Seabrook, New Hampshire, June 1986.
Technical Report No. 1589.01.
4.
Seabrook Station Emergency Planning Sensitivity Study, prepared by Pickard, Lowe and Garrick, Inc., Newport Beach, CA for New Hampshire
)
Yankee Division, Public Service Company of New Hampshire, Seabrook, 1
New Hampshire, April 1986.
PLG - 0465.
5.
John DeVincentis to George W. Knighton letter dated January 30, 1984, Seabrook Station Probabilistic Safety Assessment Main Report and Summary Report.
6.
Seabrook Station Risk Management and Emergency Planning Study, prepared by Pickard, Lowe and Garrick, Inc., Newport Beach, CA for New Hampshire Yankee Division, Public Service Company of New Hampshire, Seabrook, New Hampshire, December 1985.
PLG - 0432.
7.
Seabrook Station Probabilistic Safety Assessment (Summary Report and 6 volumes), prepared by Pickard, Lowe and Garrick, Inc., Newport Beach, CA for Public Service Company of New Hampshire, Manchester, New Hampshire l
and Yankee Atomic Electric Company, Framingham, MA, December 1983.
PLG - 3000.
8.
Seabrook Station Probabilistic Safety Assessment Technical Summary
)
Report, prepared by Pickard, Lowe and Garrick, Newport Beach, CA for
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Public Service Company of New Hampshire, Manchester, New Hampshire
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and Yankee Atomic Electric Company, Framingham, MA, June 1984.
PLG - 0365.
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1.
SEISMIC FRAGILITY UPDATE t
Seismic sequences dominate release categories S2 and S6 in the Risk Management and Emergency Planning Study (RMEPS). A separate submittal on the principal contribution to early healtn risk explains how these release categories contribute to early health risk. This write-up explains now the seismic fragility update is expected to change the frequency of S2 and S6. A complete requantification will be included in the probabilistic safety assessment (PSA) update now in progress and i
planned for completion in 1987 l
In a complete seismic risk analysis, there is first performed a point estimate analysis using the plant event trees that are quantified for l
several discrete values of ground acceleration. From tne point estimate results, dominant sequences initiated by seismic events are identified; then, these sequences are reanalyzed using a special computer code called SEIS4.
In SEIS4, the seismicity curves and fragility curves are convoluted and uncertainties in these curves are propagated to obtain uncertainty distributions on the final result, which is either a core relt or plant damage state frequency contribution.
In the following approximate analysis, the point estimate step is bypassed, so some assumptions are made about dominant sequences. Hence, these results are only rough approximations and should only be used for order-of-magnitude estimates. A complete reanalysis of seismic events is currently in progress and is planned for completion in 1987.
1.1 RELEASE CATEGORY S2 This release category is dominated by earthquake and transient initiating events. These sequences can be simply represented as OG*(DT + OG + SSPS)
(1) where OG
= Of fsite Power Fragility UT
= Diesel Generator Day Tank Fragility DG
= Diesel Generator Fragility SSPS = Solid State Protection System (SSPS) Fragility (actually 120V AC power panel required for SSPS success) and only seismic unavailabilities are included.
Also, earthquake and large loss of coolant accident (LOCA) initiating events provide a small contribution and can be represented as LL*0G*(OT + DG + SSPS)
(
1431P100186
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Equation (1) was quantified with the SEIS4 computer code and resulted in the following annual core melt frequency:
Mean
= 2.84 x 10-5 Variance = 2.24 x 10-9 Based on the fragility update, SSPS and UT can be dropped from the model, based on significantly higher capacities. However, a relay chatter fragility at a relatively lower capacity has been identified in the 4,160V switchgear. This chatter could have a negative effect; e.g., trip out the diesels. Until the consequences of this chatter are evaluated, it is assumed that the chatter fails both diesels. Therefore, Equation (1) can be changed as follows:
OG*(chatter + UG)
(3) where Chatter = Relay Chatter Fragility (4,160V switcngear)
Quantifying equations for annual core melt frequency with SEIS4 results in Mean
= 1.8 x 10-5 Variance = 9.58 x 10-10 Comparing the quantification of Equations (1) and (3) shows a slight reduction (less than a factor of 2) in frequency. However, this assumes the chatter fails the diesels without recovery. An ongoing relay chatter review will determine whether this particular chatter is a real concern.
In addition, this review will determine whether there are any other relay chatters that should be considered in the model.
1.2 RELEASE CATEGORY 56 This release category is dominated by earthquake and transient initiating events. These sequences can be simply represented as N0G*SSPS (4) where N0G = Offsite Power Available (negation of UG - fragility)
As described above under release category 52, the solid state protection system can be dropped from the model. Therefore, the simple model in Equation (4) would go to zero. To actually determine the new S6 f requency, the whole plant model needs to be requantified and unraveled to obtain new dominant sequences and frequencies. However, the trend is a reduced frequency unless the ongoing relay chatter review identifies new sequences.
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UNITED STATES
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..e Docket Nos.: 50-443 and 50-444 Mr. Robert J. Harrison President & Chief Executive Officer Public Service Company of New Hampshire Post Office Box 330 Manchester, New Hampshire 03105
Dear Mr. Harrison:
Subject:
Request for Additional Information for Seabrook Station, Units 1 and 2, Emergency Planning Sensitivity Study 1
The enclosed Request for Additional Infonnation documents the oral and handwritten questions transmitted to Public Service Company of New Hampshire personnel and contractors during our meeting in Bethesda, Maryland on September 23, 1986.
Please provide your responses promptly to facilitate our review.
Questions or additional infonnation regarding this matter should be directed to the Technical Project Manager 'for the review of the Seabrook Emergency Planning Sensitivity Study, S. M. Long (301) 492-8413.
Sincerely,
%N i
Steven M. Long, Project Manager PWR Project Directorate No. 5 Division of PWR Licensing-A
Enclosure:
As stated cc: See next page M!!9 478 E
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Enclosure REQUEST FOR ADDITIONAL INFORMATION SEABROOK STATION, UNITS 1 AND 2 DOCKET NOS.:
50-443 AND 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.
Document 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.
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 levels was not considered in the calculations (Ref. p. 35, assumption 6).
Verify that the combined effect does not change the conclusions of the analysis.
I l
l
1.
8.
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 construction problem concerning staggering of these welds, provide justification for not using a reduced ultimate strength for the rebar.
9.
The containment analysis is based on an axisymmetric 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 axisymmetric conditions and assess their impact on the failure modes and i
analysis.
i 10.
Only a sample of pipe penetrations are considered in some detail (X-23, l
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.
1
- 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 i
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 l
source term.
If this is the case, please describe all accident sequences (internal and external events) that contribute to the frequency of the S2 source term given in Table 4-2, pg. 4-7.
in particular, define how j
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.
3
- 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
{
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.
- 19. Document the effect that the recent update in seismic fragilities will have on the conclusions of the PSA results.
?
I i
1 '
- 20. Assess the impact on risk of using the assumption of ultimate containment capability predicted by VE&C analysis (150 psig).
- 21. What is the impact on risk from accidents during shutdown and refueling when the containment function may not be available?
22.
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 0.1%/ day.
Compare this assumption with the containment integrity violation a.
data presented in NUREG/CR 4220.
b.
What contributions would these containment integrity violation data make to the probabilities for each of the release categories (Assume the SSW category is redistributed over all the appropriate categories by the conditional probabilities of preexisting leakage paths of the size appropriate to each category).
4 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 shculd be specified.
l The probability of occurrence of each release category should be stated. These data should be provided for the current study and for the origieal PSA results. Changes between the two studies should be attributed to specific differences in the analysis.
b.
Provide a set of early fatality conditional probability curves for each release category, assuming evacuation distances of 1 mile and 2 miles.
l c.
Provide the conditional mean risk of early fatality for each of the curves provided in b.
r I
- 24. Provide a quantitative description of the effects of the following differences l
between the original PSA and the current study:
a.
reduction in probability of core-melt V sequences b.
factor of 1000 scrubbing of releases through RHR seals change of release category (S6 to SI) for unscrubbed event V sequences.
c.
The effects should be described in terms of differences in risk curves for early fatalities and for 200 rem vs distance.
- 25. Provide a list of all paths for loss of RCS inventory outside containment.
Show how these have been considered with respect to LOCA and with respect to containment bypass for radioactive materials following core damage.
. f 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.
28.
Identify any penetrations connected directly into the containment atmosphere which rely on any remote manual or manual valves for isolation.
l 4
l l