ML20248B625

Forwards List of Potential Discussion Items for 890727 Meeting W/Util in Rockville,Md Re Severe Accident Mitigation Issues.Meeting Previously Announced in 890724 Transmittal

ML20248B625
Person / Time
Site:	Limerick
Issue date:	07/25/1989
From:	Suh G Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8908090276
Download: ML20248B625 (7)

Text

- _ _ _ _ _ _

July 25, 1989 LDocket'Nos. 50-352/353 MEMORANDUM FOR: Distribution

( FROM: Gene Y. Suh, Project Manager Project Directorate I-2

[ Division.of Reactor Projects I/II

SUBJECT:

FORTHCOMING MEETING WITH PHILADELPHIA ELECTRIC CO. ON LIMERICK SEVERE ACCIDENT HITIGATION ISSUES The attached provides a list of potential discussion items for the subject meeting scheduled for July 27, 1989. The meeting, previously announced in a

/,uly 24, 1989 transmittal, is to be held in Room.6811, One White. Flint North, 11555 Rockville Pike, Rockville, Maryland between 1:00 PM and 5:00 PM.

'/S/

Gene Y. Suh, Project Manager Project Directorate I-2 Division of Reactor Projecti, I/II

Enclosure:

As stated cc w/ enclosure:

See next page DISTRIBUTION EDocket?FjleVi g(b[

NRC'"&~Lbcil'PDRs PDI-2 Reading I

RClark/RMartin GSuh/ETrottier cc: Licensee / Applicant & Service List

[DIST) q PDI-2/PM PDI-2/D GSuh:trd WButler 7/2f/8 g g(/89 890B090276 890725 e PDR ADOCK 05000352E P PNV g

• h6 RfG

,  ;/ %g

' ~ UNITED STATES I '!* n NUCLEAR REGULATORY COMMISSION

-(  ;$ WASHINGTON, D C. 20555

,,,,. July 25, 1989 l

Docket Nos. 50-352/353 MEMORANDUM FOR: Distribution FR0ft: Gene Y. Suh, Project Manager

. Project Directorate I-2 Division of Reactor Projects I/II

SUBJECT:

FORTHCOMIt'G MEETING WITH PHILADELPHIA ELECTRIC CO. ON LIMERICK SEVERE ACCIDENT MITIGATI0tl ISSUES The attached provides a list of potential discussion items for the subject meeting scheduled for July 27, 1989. The meeting, previously announced in a July 24, 1989 transmittal, is to be held in Room 6B11, One White Flint North, 11555 Rockville Pike, Rockville, Maryland between 1:00 PM and 5:00 PM.

, Y -

Gene Y. h, Project Manager Project Directorate I-2 Division of Reactor Projects I/II

Enclosure:

As stated cc w/ enclosure:

See next page

__ _ _ _ _ _ _ _ - - _ - _ _ _ - _ - _ - - _ _ _ _ _ _ _. _ - - _ =

Mr. George A. Hunger, Jr. Limerick Generating Station Philadelphia Electric Company Units 1 & 2-CC:

Troy B. Conner, Jr., Esquire Mr. Ted Ullrich Conner and Wetterhahn Manager - Unit 2 Startup 1747 Pennsylvania Ave., N.W. Limerick Generating Station Washington, D. C. 20006 P. O. Box A Sanatoga,' Pennsylvania 19464-Mr. Rod Krich S7-1 Philadelphia Electric Company Mr. John Doering 955 Chesterbrook Boulevard Super

• 1tendent-0perations Wayre, Pennsylvania 19087-5691 Limerick Generating Station P. O. Box A Mr. Charles W. Elliott Sanatoga, Pennsylvania 19464 Poswistilo, Elliott and Elliott 325 N.10th Street Thomas Gerusky,' Director Easton, Pennsylvania 18042 Bureau of Radiation Protection PA Dept. of Environmental Resources Mr. Graham M. Leitch, Vice President P. O. Box 2063 Limerick Generating Station Harrisburg, Pennsylvania 17120 Post Office Box A

)

Sanatoga, Pennsylvania 19464 , Single Point of Contact P. O. Sox 11880 Harrisburg, Pennsylvania 17108-1880 Mr. James Linville U.S.' Nuclear Regulatory Commission Mr. Philip .1. Duca Region I Superintendent-Technical 475 Allendale Road Limerick Generating Station King of Prussia, PA 19406 P. O. Box A Sanatoga, Pennsylvania 19464 Mr. Thomas Kenny Senior Resident Inspector US' Nuclear Regulatory Commission P. O. Box 596

• Pottstown, Pennsylvania 19464 Mr. John S. Kemper Senior.Vice President-Nuclear Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101

~

Discussion Items for July 27, 1989 Meeting with PEco

1. A comparisor of the core damage frequencies for each accident cless in the original and updated PRA, and the reasons for the changes (by accident class), including changes in models/ assumptions, hardw.tre improvements, and credit for recovery actions.

2. Views and estimates on the uncertainty associated with the core damage frequency estimates on which the PECo SAMDA assessment is based.

3. How NRC review comments / recommendations [NUREG-1068) on the original PRA were incorporated in the revised PRA, and the rationale if certain recom-mendations were not incorporated.

4 The models used to develop the release fractions for each of the accident classes / plant damage states, and to develop offsite consequence estimates.

5. How the risk reduction benefits (i.e., averted mah-rem estimates) were derived for each SAMDA, and in particular, for the core debris control devices considered. A more complete description of the techniques considered for core debris control.

n -

-4

~

e. .'

6. The. methods / assumptions used to develop the following information, and t'ne i

values used in the PRA:

a. probability and magnitude of drywell head leakage

b. probability of suppression pool bypass,.e.g., downcomer meltthrough, stuck open vacuum breaker

c. fraction of sequences in which reactor fails at'high pressure and low pressure

d. likelihood of accidents while deinerted, and containment challenges due to deinerting following venting.

7. The conditional early fatality estimates for the four dominant sequences identified in Table 1-4 of the June 23, 1989 PECo submittal. A discussion related to the changes in frequency estimates of these sequences from the original PRA.

8. A discussion related to the conservatism in the modeling of the 60minant fire sequences, particularly sequences 1, 4, 7,12, and 16 in Table 1-2 of the PECo submittal.

9. A detailed discussion regarding the reduction in tyrbine trip frequency assumed for Unit 2, and assumptions related to lots of offsite power.

10. A discussion regarding any new sequences modelled in the revised PRA, such as sequences resulting from stean line breaks outside containment.

I j

I

..s. ..;

. 11. Underlying assumptic7s and bases'for cost estimates. At a minimum, and as appropriate for each citigating' system, costs for the following cost-

.a '

elements:-

I. Initial Investment

a. new structures

b. list of hardware and materials (e.g., piping) with costs

c. labor costs - number of hours and labor rates

d. engineering / design costs

e. Quality assurance costs

f. health physics costs - occupational exposure in person rem, health physics support, etc.

g. procedural costs (e.g., Tech spec change)

h. training' costs

i. replacement energy costs - [ number of days of downtime assumed]

j. other up-front costs

11. Recurring Costs (expressed on an annual basis)

a. maintenance

b. Other recurring costs ,

12. Table 1 identifies a number of Mark 11 improvements under consideration as part of the NRC Containment Performance Improvement Program. A discussion of views on the risk reduction potentici and expected costs for each of these for Limerick.

TABLE 1 E21ential_Isst-Bansficial Severe AccidfD1 ditigation Improvements for Limerich Proposed improvements 1, 2, and 3 MUST be taken together.

Proposed improvement 4 ma'y be taken independently of the others.

1. Diesel Generator: This diesel generator is to provide power for the operation of the ADS (SRV),

controls, and for the operation instrumentation and of the RWCU system.

Susquehanna has provided a 100KW Blue Max portable generator system which can' provide power to both units via cables and splices to the existing plant. All materials and procedures to use this equipment have been provided for a total cost of

$150K. Note that the power requirement for the RWCU is not known at this time.

2. Low Pressure Backup Water Supply System: This water supply should be capab'e of injection into the reactor (when the reactor is depressurized) and used in the containment spray system. The system should have sufficient capacity to arrest core degradation once injection is initiated into the reactor vessel. Current estimates indicate that this may be  !

approximately 500 rpm. The power for this. system should be independent of the normal power systems and operable during station blackout events - such as a diesel direct drive pump  ;

or, if the diesel generator (Item 1 above) is properly sized, the pump could be powered from the Blue Max diesel.

I

3. Assured RWCU Operability: In order to prevent containment failure from use of the drywell spray system, the RWCU is needed to remove the water from the suppression pool. Thus i only the pumping system and the related valves need to be operable during a station blackout. Note that the proposed )

l diesel generator (1 tem 1 above) is supposed to be I sufficiently large to operate the pumping portion of the RWCU system. ,

4. Drywell Head Floodina: For station blaciout events or other l severe accident events where containment pressure may exceed the containment design pressure, the water from the suppression pool should be inserted, by means of a syphon,-

into the area above the drywell head and below the reactor l missile shield plugs. The drywell head bolts should be so tensioned as to leak into this pool when containment {

pressure exceeds twice design pressure but significantly below the ultimate containment failure pressure. The water will prevent over temperature failure of the drywell head seals and will provide a scrubbed release for the leakage.

It is proposed that the syphon line would be a hose that '

would be installed upon identification of the severe i accident with a hand pump to initiate flow (similar to the gasoline syphons sold by Trak Auto for $9.95).

~ _ - - _ _ - _ - _ _ - -- .--