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Summary of 851002 Meeting W/Applicant & BNL in Long Island, Ny to Discuss Nrc/Bnl Review of Probabilistic Safety Assessment,Containment Failure Modes & Radiological Source Terms.Attendance List & Draft Rept of BNL Review Encl
	ML20137A650
Person / Time
Site:	Seabrook
Issue date:	01/06/1986
From:	Nerses V; Office of Nuclear Reactor Regulation
To:	; Office of Nuclear Reactor Regulation
References
	NUDOCS 8601140439
	Download: ML20137A650 (76)
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d o*^ UNITED STATES '

8" n NUCLEAR REGULATORY COMMISSION

%. <E WASHING TON, D. C. 20555 g..... JM o s 1986 .

Docket Nos.: 50-443 and 50-444 APPLICANT: Public Service Company of New Hampshire FACILITY: Seabrook Station, Unit 1 and 2

SUBJECT:

MEETING

SUMMARY

On October 2,1985, Nuclear Regulatory Commission, Brookhaven National Laboratory and applicant representatives met at Brookhaven, (Long Island, N.Y.) to discuss the NRC/BNL review of the Seabrook Station probabilistic safety assessment (in particular containment failure modes and radiological source terms). Enclosure 1 is the list of attendees. Enclosure 2 is the draft report of the BNL review which formed the basis of the discussion in this meeting.

A number of applicant comments on specific items in the draft report were made and subsequently documented. These are given below. The comment numbers correspond to report sections identified with the'same number:

(1) Page 5: Other key parameters to include in this table are:

a. ' Containment design - reinforced. '

b. Steam generator secondary inventory - 112,000 lbm/ steam generator.

(2) Page 21: After the vessel blowdown transient to about 70-80 psia, the RCS pressure builds up rather slowly.

(3) Page 22 and 23: The SSPSA used 0.1%/ day as the design basis leak rate per Amendment 47.

(4) Page 23 and 24: The SSPSA used 100% in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months as the boundary between Type A and Type B containment leaks. Anything greater would be treated as a single puff release. This is conservative compared to the BNL definition of 100% in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

(5) Page 28: The Type A leak area of 0.5 square inches per pipe is based on failure of the weld at the worst and probably weakest location in the annular seal plate.

060114o439og%43 PDR ADOCK O PDR A

(6) Pages 28, 29 and 31: Simultaneous failure of all 4 feedwater lines is unlikely given the uncertainties in the failure pressure. After failure of the first feedwater pipe, the leak area would be large enough to rer"It in a steady or decreasing containment pressure. This would prevent failure of the other feedwater pipes. Even if all four feedwater pipes were to fail, the combined leak area would not be much larger than that for. single pipe failure. The failure mode is self-regulatino and the total leak area is determined by the pressurization source, i.e., the steaming rate or the gas generation rate.

(7) Page 30: Failure of the outer purge isolation valve and the electrical penetrations is very unlikely for the same reason. The heat losses into the concrete are too large to raise the temperature in the penetration before mass transfer into the interspace stops. Steam condensation leads to a nonconvertible atmosphere in the interspace which is in pressure equilibrium with the containment. From this point on, only conduction through the sleeve can transfer heat into the penetration.

(8) Page 30: Electrical penetration inteority is based on:

a. Location of the plugs on the outside of the penetration.

b. Calculated heat losses to the concrete. (See comment 7).

(9) Page 31: Use of Figure 3.10 would not have changed any of the SSPSA results.

(10) Page 31: The SSPSA took no credit for the secondary enclosure building except for intact containment leakage. Even type A ar.d B leaks develop at a pressure where interference at the' equipment hatch is expected to fail the enclosure locally.

(11) Page 33: Impaired evacuation was modeled for the earthquake scenarios, assuming larger delay times and slower evacuation speeds.

(12) Page 38: For top event 12 on the containment event tree, both paths are containment failure. Success is small leakage and failure is cross leakage.

(13) Page 42: Release Category S2 only represents sequences where the leak rate increases substantially (to 40%/ day) at vessel melt-through. It does not represent failure of the Type A oenetrations at 181 psia during slow pressurization sequences. Release Category S2 was introduced ,

because a PWR1 or PWR2 type release at Seabrook is extremely unlikely.

The probability of accident sequences with an isolated containment resulting in an 51 or S2 release category was based on an uncertainty analysis for the pressure spike at vessel melt-through compared to the containment pressure capacity; i.e.,

Pr(S2)=Pr(PatVMT*181 psia)

Pr(SI) = Pr (P at VMTr 211 psia)

Plant damage states for steam generator tube rupture sequences were assigned to the most appropriate release category without an attempt to compute SGTR specific release categories.

(14) Pace 51: The energy release of 3 x 109 Btu /hr only applies to gross containment failures and resulted from two considerations, namely:

1. Mechanistic failure propagation for containment membrane failure modes.

2. An eneray release which limits the plume rise to the inversion layer.

The latter limitation (2) was found more constraining.

(15)Page52: The start of the second puff release coincides with the beginning of the vaporizotion release. The duration of the vaporization release is only 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months but the duration of the second puff is 7.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to better balance the release fractions associated with each puff. In any event, the entire release for the second puff occurs at the beginning of the second puff. Therefore, the duration is only of secondary importance.

(16) Page 52: The applicant feels it would be appropriate to point out that the BNL recommendation for reducina release times and durations, as well as the conversion to single puff releases is only for the purpose of bringing the release category parameters within the CRAC code limitations and not to make the release more realistic. The applicant does not believe that the RSS times are more reasonable.

(17) Page 56: To the applicant the single puff release concept is not a realistic one because the applicant stated that they now know that all the risk significant release at Seabrook extend over several hours., The purpose of an "eouivalent" single ouff release can only be tn either assess the conservatism of a single puff release versus a multipuff release or to bring the analysis within the constraints of the CRAC code.

In either case, this should be clearly stated.

(18) Pages 52 and 56: Table 4.7 lists BNL recommended release characteristics which are based on the MARCH / CORRAL calculated point estimates, without taking into consideration any of the advances since WASH-1400. In the SSPSA source term uncertainly analysis, it is shown that these release characteristics represents something like a 99% confidence level for nonexceedance. The applicant thinks it could be appropriate to list in a ',

separate table the single puff equivalent of the SSPSA best estimate release categories. These would be 53-d, 53V-d, (53V1-d + 53V2-d +

5BV3-d) and (52V1-d + 52V2-d + SIV3-d). These could be listed simply for comparisen without a recommendation, to give the future user a choice of using whatever he believes to be appropriate for his purpose.

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JRN 9 6 W (19) Pages iii and 60: The applicant believes that the followino three conclusions are equally significant:

Local containment failures with self-reculating leak areas are shown to occur before gross containment failure. These result in extended releases with reduced consequences.

The timing of overpressure failure is measured in days compared to hours in WASH-1400.

An analysis of the source term uncertainties has shown that the best estimate source terms can be expected to be between one and two orders of magnitude lower than the point estimate releases calculated for conservative accident sequences using the WASH-1400 methodology.

(20) Page 52: The point estimate release categories were determined on the basis of a containment capacity corresponding to a wet containment condition. For the dominant release cateaories which are all dry containment conditinrs, the release timos and release fractions were corrected to the dry conditions and are shown in the uncertainty analysis. Therefore, release cateqory T)V-a on Table 11.6-16 should be used for this comparison. This release cateoorv (T/V-a) reflects the release timino calculated for this tvoe of accident sequence using the WASH-1400 methodology.

(s)

Victor Nerses, Project Mananer PWR Pro.iect Directorate No. 5 Division of PWR Licensing-A

Enclosures:

As stated cc: See next page es/yt 1/b/86 l

Meeting Summary Distribution Docket'or Central File NRC Participants NRC PDR Local PDR W. Lyon PD#5 Reading File J. Partlow (Emergency Preparedness only)

Branch Chief Projet.t Manager OELD E. Jordan

8. Grimes ACRS(10) cc: Licensee and Plant Service List e

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l (19) Paoes 111 and 60: The applicant believes that the following three

! conclusions are equally significant:

i Local containment failures with self-regulating leak areas are shown l to occur before gross containment failure. These result in extended j releases with reduced consequences.

l The timina of overpressure failure is measured in days compared to hours in WASH-1400.

l An analysis of the source term uncertainties has shown that the best estimate source terms can be expected to be between one and two orders of magnitude lower than the point estimate releases calculated for conservative accident sequences using the WASH-1400 methodology.

(20) Page 52: The point estimate release categories were determined on the basis of a containment capacity corresponding to a wet containment condition. For the dominant release categories which are all dry containment conditions, the release times and release fractions were corrected to the dry conditions and arej hown in the uncertainty analysis. Therefore, release category R V-a on Table 11.6-16 should be used for this comparison. This release cateaory (IIV-a) reflects the release timina calculated for this type of accident sequence using the WASH-1400 methodology.

h Victor Nerses, Project Manager PWR Proiect Directorate No. 5 Division of PWR Licensing-A

Enclosures:

As stated -

cc: See next page e

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Mr. Robert J. Harrison Public Service Company of New Hampshire Seabrook Nuclear Power Station ,

cc:

Thomas Dignan, Esq. E. Tupper Kinder, Esq.

John A. Ritscher, Esq. G. Dana Bisbee, Esq.

Ropes and Gray Assistant Attorney General 225 Franklin Street Office of Attorney General Boston, Massachusetts 02110 208 State Hosue Annex Concord, New Hampshire 03301 Mr. Bruce B. Beckley, Project Manager Public Service Company of New Hampshire Resident Inspector Post Office Box 330 Seabrook Nuclear Power Station Manchester, New Hampshire 03105 c/o US Nuclear Regulatory Comission Post Office Box 700 Dr. Mauray Tye, President Seabrook, New Hampshire 03874 Sun Valley Association 209 Sumer Street Mr. John DeVincentis, Director Haverhill, Massachusetts 01839 Engineering and Licensing Yankee Atomic Electric Company Robert A. Backus, Esq. 1671 Worchester Road O'Neil, Backus and Spielman Framingham, Massachusetts C1701 116 Lowell Street Manchester, New Hampshire 03105 Mr. A. M. Ebner, Pro.iect Manager United Engineers & Constructors Ms. Beverly A. Hollingworth 30 South 17th Street 7 A Street Post Office P.ox 8223 ;

Hampton Beach, New Hampshire 03842 Philadelphia, Pennsylvania 19101 William S. Jordan III Mr. Philip Ahrens, Esq.

Diane Curran Assistant Attorney General Harmon, Weiss & Jordan State House, Station #6' 20001 5 Street, NW Augusta, Maine 04333 Suite 430 Washington, DC 20009 Mr. Warren Hall Jo Ann Shotwell, Esq. Public Service Company of Office of the Assistant Attorney General New Hampshirr Environmental Protection Division Post Office Box 330 One Ashburton Place Seabrook, New Hampshire 03874 Boston, Massachusetts 02108 Seacoast Anti-Pollution League D. Pierre G. Cameron, Jr., Esq. Ms. Jane Douchty General Counsel 5 Market Street Public Service Company of New Hampshire

Portsmouth, New Hampshire 03801 ,

Post Office Box 330 Manchester, New Hampshire 03105 Mr. Diana P. Randall 70 Collins Street Regional Administrator, Region I Seabrook, New Hampshire 03874 U.S. Nuclear Regulatory Comission 631 Park Avenue Richard Hampe, Esq.

King of Prussia, Pennsylvania 19406 New Hampshire Civil Defense Agency 107 Pleasant Street Concord, New Hampshire 03301

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Public Service Company of Seabrook Nuclear Power Station New Hampshire ,

cc:

Mr. Calvin A. Canney, City Manager Mr. Alfred V. Sargent.

City Hall Chairman 126 Daniel Street Board of Selectmen Portsmouth, New Hampshire 03801 Town of Salisbury, MA 01950 Ms. Letty Hett Senator Gordon J. Humphrey Town of Brentwood ATTN: Tom Burack RFD Dalton Road U.S. Senate Brentwood, New Hampshire 03833 Washington, D.C. 20510 Ms. Roberta C. Pevear Senator Gordan J. Humphrey Town of Hampton Falls, New Hampshire ATTN: Herb Boynton Drinkwater Road 1 Pillsbury Street Hampton Falls, New Hampshire 03844 Concord, New Hampshire 03301 Ms. Sandra Gavutis Mr. Owen B. Durgin, Chairman Town of Kensington, New Hampshire Durham Board of Selectmen RDF 1 Town of Durham East Kingston, New Hampshire 03827 Durham, New Hampshire 03824 Charles Cross, Esq.

Chairman, Board of Selectmen Shaines, Mardrigan and Town Hall McEaschern South Hampton, New Hampshire 03827 25 Maplewood Avenue Post Office Box 366 Mr. Angie Machiros, Chairman Portsmouth, New Hampshire 03801 Board of Selectmen

for the Town of Newbury .

Newbury, Massachusetts 01950 Mr. Guy Chichester, Chaiman Rye Nuclear Intervention Ms. Cashman, Chairman Committee Board of Selectmen c/o Rye Town Hall Town of Amesbury 10 Central Road Town Hall -

Rye, New Hampshire 03870 Amesbury, Massachusetts 01913 Jane Spector Honorable Richard E. Sullivan Federal Energy Regulatory Mayor, City of Newburyport Commission Office of the Mayor 825 North Capital Street, NE City Hall Room 8105 Newburyport, Massachusetts 01950 Washington, D. C. 20426 ,

Mr. Donald E. Chick, Town Manager Mr. R. Sweeney Town of Exeter New Hampshire Yankee Division 10 Front Street Public Service of New Hampshire Exeter, New Hampshire 03823 Company 7910 Woodmont Avenue Mr. William B. Derrickson Bethesda, Maryland 20814 Senior Vice President Public Service Company of New Hampshire Post Office Box 700, Route 1 Seabrook, New Hampshire 03874

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Enclosure 1 Attendees to the Octcher 2, 1985 Seabronk Meeting NRC W. Lyon BNL M. Khatib-Rahbar S PSNH A. Torri M. Moody K. Kiper R. Sweeney

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'* Enclosura 2 l i

, b e. BNL/NUREG-NUREG/CR-ki l ; -

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A REVIEW OF THE SEABROOK STATION PROBABILISTIC SAFETY ASSESSMENT: CONTAINMENT FAILURE MODES AND RADIOLOGICAL SOURCE TERMS M. Khatib-Rahbar A. K. Agrawal, H. Ludewig and W. T. Pratt .

September 1985 -

Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 4

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' tit ABSTRACT e

A technical review and evaluation of the Seabrook Station Probabilistic.

Safety Assessment has been performed. It is determined that (1) containment

~ response to severe core melt accidents is judged to be an important factor in -

mitigating the consequences, (2) there is negligible probability of prompt containment failure or failure to isolate, (3) failbre during the first few

hours after core melt is also unlikel (4) the point-estimate radiological releases are comparable in magnitude t those used in WASH-1400, and (5) the energy of release is somewhat higher t an for the previously reviewed studies.

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-v-CONTENTS .

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ABSTRACT................................................................ iv ACKNOWLEDGMENT.......................................................... vi .

LIST OF TABLES.......................................................... vii

LIST 0F FIGURES.........................................................

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1. INTR 000CTION.......................................................

1 1.1 Background....................................................

1 1.2 Obj ec ti v e s a n d Sc o pe . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 1.3 Organization of the Report....................................

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2. PLANT DESIGN AND FEATURES IMPORTANT TO SEVERE ACCIDENT ANALYSIS....

Design.................................... '2 -

,- 2.1 Assessment of Plant 4 2.2 Compa ri so n wi th Othe r P1 ants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3. ASSESSMENT OF CONTAINMENT PERFORMANCE.............................. 7 ,

7 3.1 Background....................................................

8 3.2 Co n ta i nment Fa il u r e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3.2.1 Background..........................................'... 8 8

3.2.2 Design Description.....................................

21 3.2.3 Leakage Rate Calculation...............................

22 3.2.4 Contai nment Fail ure Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.2.4.1 Lea k-Be fo re-Fail ure . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.2.4.2 Cl assi ficati on o f Fail ure. . . . . . . . . . . . . . . . . . . . . 24 4 3.2.5 Containment Pressure Capacity.......................... 24 3.2.5.1 Concrete Containment.......................... 27 3.2.5.2 Liner......................................... 27 /

3.2.5.3 Penetrations.................................. 31

' 3.2.5.4 Containment Fail ure Probabil ity. . . . . . . . . . . . . . . '

31 3.2.5.5 Containment Encl os ure . . . . . . . . . . . . . . . . . . . . . . . . .

3.3 Definition of Plant Damage States and Containment Response C1 asses.............................................. 31

' 3.4 Containment Event Tree and Accident Phenomenology............. 33 3.5 Containment Matrix (C-Matrix)................................. 38 l 1

3.6 Rel ea se Category Frequenci es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44

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4. ACCIDENT SOURCE TERMS..............................................

i 48 l 4.1 Asse ssment of Severe Accident Sou rce Te rms . . . . . . . . . . . . . . . . . . . .

4.2 Sou rce Te nn Unce rta i n ty An al y si s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 ~

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................... 56 i 4.3 Re commend ed So u rc e Te rms . . . . . . . . . . . . . . . . .

5.

SUMMARY

AND CONCLUSIONS............................................

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6. REFERENCES......................................................... ,

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-vi-LIST OF TABLES Table Title Page 2.1 Compari son of Sel ected Design Characte ri stics. . . . . . . . . . . . . . .. . . . . 5 3.1 Containment Operating and De si gn Pa ramete rs. . . . . . . . . . . . . . . . . . . . . . 10 3.2 Cont ai nment Li ne r Pe ne t ra ti ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 1

I 3.3 Leak Area Estimates fc Mechanical Penetrations... ... . ... . .. . . . . . . 29 3.4 Frequencies of Occurrence of the Plant Damage States............. 35 3.5 Contai r. ment Re sponse Cl ass De fi ni tions . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 3.6 Co ntainment Cl as s Mean Frequenci e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 3.7 Accident Fnase and Top Events for the Seabrook Containment Event Tree....................................................... 39 3.8 Release Categories Employed in the Seabrook Station Risk Mode 1............................................................ 40 3.9 Simpl i fi ed Containment Matrix for Se ab rook. . . . . . . . . . . . . . . . . . . . . . . 41 3.10 Frequency of Dominant Release Categories (yr-1)..... . ..... . . . . . . . 45 3.11 Contribution of Containment Response Classes to the Total Core Melt Frequency.............................................. 46 3.12 Release Category Frequency as a Fraction of Core Melt Frequency........................................................ 47 ,

4.1 Seabrook PointhEstimate Rel ease Categories...... .. . . . . . . . . . . . . . . . 49 s

4.2 La te Overpre s suri zation Fail ure Compari son. . . . . . . . '. . . . . . . . . . . . . . . 51 4.3 Comparison of Releases for Failure to Isolate Containment a nd the By-Pa s s Se que nc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53

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4.4 Com pa r i son o f .AB- c a nd TMLB '- c ( BM I-2104 ) to 137 and llf. . . . . . . . . . 55 4.5 Compa ri son of 1RDT (sum) to V-sequence (Surry) . . . . . . . . . . . . . . . . . . . . 57 4.6 B NL - S u g g e s t e d So u rc e Te rm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 5 8 C

4.7 BNL-Suggested Release Characteristics for Seabrook.............../ 59 t

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, LIST OF FIGURES .

Figure

Tit 1e Page 3.1 A schematic represenution of source term calculation 5........... 9

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. Equipment hatch with personnel air 1cck............................

3.2 12 3.3 Pe r s o n n el a i rl o c k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.4 Typi cal hi gh energy pi pi ng penetrati on. . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.5 Typi cal moderate energy pi ping penetrati on. . . . . . . . . . . . . . . . . . . . . . . 16 3.6 Typi cal el egtri :a1 penetration. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 1

3.7 Typi cal ventil ati on pene tration. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.8 A pictorial representation of laahage categories................. 25 3.9 Estimated radial displacement of containment wa11................ 26 3.10 Estimated containment failure fractions.......................... 32 3.11 Definitions of the plant damace states used in SPSS.......... .... 34 e

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1. INTRODUCTION 1.1 Ba:kground_ - ;

Probabilistic Risk Asessment (PRA) studies have been undertaken by a num-ber of utilities (as exemplified by Refs.1-4) and submitted to the Nuclear '

Regulatory Commission (NRC) for review. Brookhaven National Laboratory (BNL) y under contract to the NRC, has been involved in reviewing core malt phenome-i nology, containment response and site consequence aspects of the PRAs.

This report presents a review and evaluation of the containment failure modes and the radiological release characteristics of the Seabrook Station Probabilistic Safety Assessment (SSPSA), which was completed by Pickard, Lowe and Garrick, Inc. (PLG) for the Public Service Company of New Hampshire and l

Yankee Atomic Electric Company in December 1983.5 !

j i 1.2 Objective and Scope The objective of this report is to provide a perspective on severe acci-dent p opagation, contaiment response and failure modes together with radiol-ogical source term characteristics for the Seabrook Station. Accident initia-tion and propagation into core damage and meltdown sequence; nre reviewed by -

' the Lawrence Liverme,re National Laboratory (LLNL) as repor:'n 'n an incomplete report [6] prepared for the Reliability and Risk Assessment f.,anch of NRC.

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? In the pres report, principal contaiment design feaa.res are dis-ose of Zion, Indian Point and Millst.me-3 designs.

cussed and compared w th Those portions of th PSS Telated to severe accident phenomena, containment l respcnse ar.d radio g source terms are described and evaluited. t'umerical 1 7 I,

adjustments to the SPS / estimates are documented and justified.

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'i 1.3 Organization of t'e Report At brief review p/f the Seabrook plant features important to severe acci dent analysis is preyented in Chapter 2 along with comparisons to Zion, Indian Point and Millstone 3 plant designs. Chapter 3 contains the assessment of containment perform nce. Specifically, definition of contalment response classes and plant d age states, analytical ho containment failure model,

contaiment event ree and accident phenomeno - y and the cont.inment matrix are reviewed. Ch ter 4 addresses the accide t source terms together with l

justifications for adjustment where necessary. The results of this review are sumarized in Chap' er 5. .

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2. PLANT DESIGN AND FEATURES IMPORTANT TO SEVERE ACCIDENT ANALYSIS In this section, those piant design features that may be important to an assessment of degraded and core melt scenarios and contaiment taalysis are reviewed. These important features are then ccnpared with the lion, Indian Point and Millstone-3 facilities to identify comonalities for benchmark ,

comparisons.

2.1 Assessment of Plant Design The Seabrook Station is comprised of two nuclear units each having an identical Nuclear Steam Supply System (NSSS) and turbine generator. The units are arranged using a " sling-along" concept which results in Unit 2 being arranged similar to Unit 1 but moved some 500 feet west. Each unit is a 1150 MWe (3650 MWt), 4-loop Westirthouse PWR plant. The turbine-generators are supplied by the General Elactric Company and the balance of the plant is designed by United Engineers and Constructors.

Each containment completely encloses an NSSS, and is a seismic Category I reinforced concrete structure in the form cf a right vertical cylinder with a hemispherical top doma and flat foundation mat built on bedrock. The inside face is lined with a welded carbon steel plate, providing a high degree of .

leak tightness. A protective 4 ft. thick concrete mat, which forms the floor of the contai ment, protects the liner over the foundation mat. The containment structure provides biological shielding for nomal and accident conditions. The approximate dimensions of the contaiment are:

Inside diameter 140 ft. -

Inside height 219 ft.

Vertical wall tt.!ckness 4 ft. 6 in and 4 ft. 71/2 in.

Dome tnickness 3 ft. 6 1/8 in. '

Foundation mat thickness 10 ft.

Contalment penetrations are provided in the lower portion of the structure.

and consist of a personnel lock and an equipment hatch / personnel lock, a fuel transfer tube, electrical, instrumentation, and ventilation penetrations.

Each contaiment enclosure (also known as secondary contaiment) su r-rounds a containment and,is designed in a similar configuration as a vertical ,

right cylindrical seismic Category I, reinforced concrete structure with dome and ring base. The approximate dimensions of the structure are: inside diam-eter,158 ft; vertical wall thickness, varies from 1 ft, 3 in. .to 3 ft; and dome thickness,1 f t, 3 in.

The containment enclosure is designed to collec any leakage from the

/)( contaiment structure other than leakage associated w th piping, electrical and access passage penetration and discharge to the filtration system of .

t contai nment. To accomplish this, the space betweerr the contaiment enclosure and the containment structure, as well as the penetration and safeguards jump.

areas, are maintained at a negative pressure following a design basis accident by fans which take suction from the containment enclosure and exhaust to atmosphere through charcoal filters. To ersure air tightness for the negative, pressure, leakage through all joints and penetrations has been minimized.

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A containment spray system is utilized for post accident containment heat i renoval. The containment spray system is designed to spray water containing boron and sodium hydroxide int'o the containment atmosphere after,a major acci-dent to cool it and remove iodine. The pumps initially take suction from the refueling water storage tank and deliver water to the containment atmosphere '

through the spray headers located in the contaiment dome. After a prescribed amount of water is removed from the tank, the pump suction is transferred to the contalment sump, and cooling is continued by recirculating sump water j through the spray heat exchangers and back through the spray headers.

The spray is actuated by a containment spray actuation signal which is 4 generated at a designated contaiment pressure. The system is completely re-dunda.it and is designed to withstand any single failure.

The con:ainment isolation system establishes and/or maintains isolation of the contaiment .from the outside enviroment in order to prevent the re-lease of fission products. Automatic trip isolation signals actuate the ap-j propriate valves to a closed position whenever automatic safety injection oc-cers or high contaiment pressure is experienced. Low capacity thermal elec-tric hydrogen recombiners are provided.

The emergency core cooling system (ECCS) injects borated water into the .

l reactor coolant system following accidents to limit core damage, metal-water reactions and fission product release, and to assure adaquate shutdown mar-gin. The ECCS also provides continuous long-term post-ac ident cooling of the core by recirculating borated water between the contalmenc sump and the reac-tor core.

i The ECCS consists of two centrifugal charging pumps, two high pressure safety injection pumps, two residual heat removal pumps and heat exchangers, and four safety injection accumulators. The system is completely redundant, and will assure flow to the core in the event of any single failure.

4 The control building contains the building services necessary for contin-

] uous occupancy of the control room complex by operating personnel during all l operating conditions. These building services include: HVAC services, air

} purification and iodine removal, fresh air intakes, fire protection, emergency i breathing apparatus, communications and meteorological equipment, lighting, i and housekeeping facilities.

l Engineered Safety Feature (ESF) filter systems required to perform a .

i safety-related function following a design basis accident are discussed below:

a. The contaiment enclosure exhaust filter system for each unit col-lects, filters and discharges any containment leakage. The system is not normally in operation, but in the event of an accident, ittis

placed in operation and keeps the containment enclosure and the building volumes associated with the penetratien tunnel and the .ESF ;

equipment cubicles under negative pressure to ensure all leakage f ront.

l the contalment structure is collected and filtered before discharge to the plant vent.

1 l

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b. One of two redundant' charcoal filter exhaust trains is placed in operation in the fuel

storage building whenever irradiated fuel not in a cask is being handled. These filter units together wi.th dampers.

and controls will maintain the building at a negative pressure.

4

! The emergency feedwater system supplies demineralized water from the con-

densate' water storage tank to the four steam generators upon loss of nomal feedwater flow to remove heat from the reactor coolant system. Operation of l

i i the system will continue until the reactor coolant system pressure is reduced ,

l te a value at which the residual heat removal system can be operated. The

! combination of one turbine-driven and one motor-driven emergency feedwater pump provides a diversity of power so;rces to assure delivery of condensate

under emergency conditions.

The two units of the facility are interconnected to off-site power via 4

three 345 kilovolt' lines of the transmission system for the New England i states. The nomri preferred source of pcwer for each unit is its own main turbine generator. Tae redundant safety feature buses of each unit are power- .

ed by two unit auxiliary transfomers. A highly reliable generator breaker is l

provided to isolate the generator from the unit auxiliary transfomers in the

) event of a generator trip, thereby obviating the need for a bus transfer upon '

loss of turbine generator power. In the event that the unit auxiliary trans-fomers are not available, the redundant safety feature buses of each unit are powered by two reserve auxiliary transfomers. Upon loss of off-site power, i

i each unit is supplied with adequate power by either of two fastestarting, diesel-engine generators. Either diesel-engina generator ar.d its associated safety feature bus is capable of providing acequate power for a safe shutdown i'

. under accident conditions with a concurrent loss of off-site. power. A con-i stant supply of power to vital instruments and controls of each unit is assur-ed through the redundant 125 volt direct current buses and their associated

battery banks, battery chargers and inverters.

2.2 Comparison with Other Plants .

j Table 2.1 sets forth the design characteristics of the Zion, Indian'

' Point-2, and Millstone-3 facilities as they compare to the Seabrook station. ,

i l

It is se'en that the contairsnent. characteristics are quite similar with i the exception of containment operating pressure for Millstone-3 (subatmospher-

' ic design), and the use of fan coolers in Zion and Indian Point for post-acci-dent containment cooling, the lower reactor cavity configuration, and chemical composition of the concrete mix. The primary system designs are nearly iden-tical between the four units.

The Seabrook containment building basemat and the internal concrete structures are composed of basaltic-based concrete. As concrete is heated, water vapor and other gases are released. The initial gas consists largely of carbon dioxide, the quantity of which depends on tr.e amount of calcium caibon-ate in.the concrete mix. Limestone concrete can contain up to 80% calcium.

carbonate by weight, which could yield up to 53 lb of carbon dioxide per cubic foot of concrete. However, basaltic-based concrete contains very little cal- '

cium carbonate (3.43 w% for Seabrook) and would not release a substantial amount of carbon dioxide.5 Thus, pressurization of the containment as a l

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Table 2.1 Comparison of Selected Design Characteristics Zion Indian Point Millstone Seabrook Unit 13 Unit 2 3 Unit 3".7 Unit 1,2 6 D2 sign Parameters 3,250 3,030 3,411 3,650 Rsactor Power (MW(t)]

Containment Building: (o (ft )3 2.73 x 10 6 2.61 x 10 6 2.3 x 10 6 2.7 x 10 6 Free Volume 6 .7 Desigr Pressure (psia) 62 62 59.7 (psia) 14.7 12.7/9.1 15.2 Initial Pressure 15 120 120/80 120 Initial Temperature (*F) 120 Primary System:

(ft 33) 12,710 11.347 11,671 13 140,7N Water Volume @ l2r Steam Volume (ft ) 720 720 216,600 7

222,739 222,739 (it.) 216,600 Mass of U02 in Core 21,000 20,407 7 19,000 Mass of Steel in Core (1b) 45,234 44,500 44,600 45,296 Mass of Zr in Core (1b) 87,000 87,000 Mass of Bottom Head (1b) 87,000 78,130 14.4 14.4 B3ttom Hatd Diameter (ft) 14.4 14.7 0.45 0.45 Bsttom Head Thickness (ft) 0.45 0.44 Containment Building Coolers:

yes yes yes yes Sprays no Fans (with safety function) yes yes no Accumulator Tanks:

200,000 173,000 348,000 213,000 Tctal Mass of Water (1b) - 615 Initial Pressure (psia) 665 665 600 T;mperature _ (

F ) 1, 150 150 , 80 $ /00-/O R; fueling Water Storace Tank:

Tctal Mass of Water (1bj . 2.89 x 10 6 2.89 x 10 6 107 2.89 x 106

120 50 86 Temperature ('F) 100 Rnactor Cavity:

Wet Dry

Dry / Wet Configuration Wet Basaltic Concrete Material Limestone Basaltic Basaltic (Minimum (Maximum Capacity = 3.9 x 10 6 lb)

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result of corium/ concrete interactions would be expected to take a very long . I time.

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3. ASSESSMENT OF CONTAINNENT PERFORMANCE In this chapter, the review of contaiment responce to severe accidents is described. Analytical techniques used to analyze core meltdown phenomena.

and containment response are reviewed, containment failure model is assessed and pl ant damage states and contai nment failure modes Finally, are evaluated.

Parallels between this study and other PRAs are cet forth. the rel-evance and validity of the conclusions is addressed.

3.1 Contaiment Analysis Methods A brief description of the computer codes used to perform the transient degraded, core maltdown and containment response analyses is provided in this section.

The MARCHs computer code is used to model the core and primary system transient behavior and to obtain mass and energy releases from the primary system until reactor vessel failure. These mass and energy releases are then used as input to tne other computer codes for analysis of containment re-sponse.

For sequences in 'which the reactor coolant system remains at an elevated S .

pressure until the vessel failure (" time-phased dispersal"), the MOOMESH computer code is used. This code calculates the steam and hydrogen blowdown from the reactor vessel using an isothermal ideal gas model. The water level boil-off from the reactor cavity 'loor is modeled using a saturated critical heat flux correlation. AdditionUly, the accumulator discharge following de-pressurization caused by the vessel failure is also considered.

S code is used to replace the INTERa sub-A modified version of the CORCON routine of the MARCH code. CORCON models the core-concrete interaction after the occurrence of dryout in the reactor cavity. The mass and energy releases from the core-concrete interaction are transferred to the M00 MESH code for proper sequencing and integration into the overall mass and energy input to COC0 CLASS 9 5 code.

i COC0 CLASS 9, a modified version of the Westinghouse COCO computer code l

utilizes the mass and energy inputs to the contaiment as computed by MARCH to 1

model the containment building pressurization and hydrogen combustion phenom-ena. This code replaces the MACE subroutine of the MARCH code.' The code also models heat transfer to the containment structures and capability for contain-

~

I ment heat removal through containment sprays and sump recirculation.

Fission product transport and consequence calculations are performed using the CORRAL-Il and the PLG proprietary CRACIT S

computer codes, respec-tively.

The analytical methods ~ used to carry out the enre and contaiment therpal hydraulics, and fission product transport calculations are identical to th'ose l l

used for MPSS-3.7 l

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-8 3.2 Containment Failure 3.2.1 Eackground '

In order to assess the risk of the Seabrook-1 plant, radiological source tems have to be calculated. Many steps are involved in such calculations.

'These are schematically shown in Fig. 3.1. The mode and time of containment ,

failure directly impact on the radioactivity release categories. These, when coupled with the status of reactor cavity and the spray system, determina the source terns. This section deals with the mode and time of containment fail-ure.

3.2.2 Design Description The primary containnent of the Seabrook plant is a seismic Category I re-inforced concrete dry structure. It consists of an upright cylinder topped with a heaispherical dome. The inside diameter of the cylinder is 140 feet and the 1:, side height from the top of the basemat to the apex of the dome is approximately 219 feet. The cylindrical vall is 4*6" thick above elevation 5' and 4'7-1/2 " thick below that evaluation. 'The dome is 3'6-1/8" thick and 69'11-7/8" in radius. The cylinder is thickened to provide room for addition-al reinforcing steel around the openings for~ the equipment hatch and the per ,

sonnel airlock. The net free volume of the containment is approximately 2.7 x 108 ft 3. ya The inside of the containment isjwelded yCt;h?4 steel liner. The' liner plate i the cylinder is 3/8" thick in all areas except penetration and the n tion of the basemat and cylinder where it is 3/4" thick. This liner

/ serves as a leak-tight membrane. Welds that are embedded in the concrete and not readily accessible are covered by a leak chase system which permits leak testing of these welds throughout the life of the plant. The dome liner is 1/2" thick and flush with the outside face of the cylindrical liner. The operating and the design parameters of containment are noted in Table 3.1.

The containment building is surrounded by an enclosure. The contairenent enclosure is a reinforced concrete cylindrical structure with a hemispherical dome. The inside diameter of the cylinder is 158 feet. The vertical wall varies in thickness from 36 inches to 15 inches; the dome is 15 inches thick.

The inside of the dome is 5'5" above the top of the containment dome. Located at the outside of the enclosure building 'is the plant vent stack, consisting of a light steel frame with steel plates varying in cross-section. The stack carries exhaust air from various buildings. '

The containment enclosure is designed to control any leakage f rom the containment structure. To accomplish this, the space between the containment and the enclosure building (approximately 4'6" wide) is maintained at a slight negative pressure (-0.25" water gauge) duri ng accident conditions by fans "

which -take suction f rom the containment enclosure and exhaust to atmosphere through charcoal filters. 7 There are a number of containment penetrations which are steel components ,

that resist pressure. These penetrations are not backed by structural con- '

l crete and include.the following:

i I

9 CONTAINMENT . TIME OF FAILURE FAILURE MODE .

WET OR ORY RELEASE SPRAY REACTOR CAVITY CATEGORY SYSTEM SOURCE TERM C

Figure 3.1 A schematic representation of source tem calculation. .

Table 3.1 Containment Operating and Design Parameters

~

Parameter . Value Normal-Operation -

Pressure , psig- 0.5 Inside Temperature , F 120 Outside Temperature , F 90 Relative Humidity , %

45 Service Water Temperature , F 80 Refueling Water Temperature , F 86 Spray Water Temperature F 88 ,

Containment Enclosure Pressu.e , inches w.g. -0.25 Design Conditions Pressure , psig 5'2.0 Temperature , F' 296 Free Volume , ft3 2.7x106 Leak Rate , 5 mass / day 0.[

Containment Enclosure Pressure , psig -3.5 o C, l z

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1. Equipment hatch,

2. Personnel air lock,

3. Piping penetrations, *

4. Electrical penetrations. .

5. Fuel transfer tube assembly,

6. Instrumentation penetrations, and 7.. Ventilation penetrations.

These components penetrate the containment and containnent enclosure shells to provide access, ar.chor piping, or furnish some other opera. tonal requirement.

All penetrations are anchored to sleeves (or to barrels) which are embedded in the concrote containment wall.

Equipment Hatch The equipment . hatch (Fig. 3.2) consists of the barrel, theThe spherical center-dished cover plate with flange, and the air lock mounting sleeve. The line of the hatch is located at elevation 37'1/2" and an azimuth of 150*.

hatch opening has an inside diameter of 27'5". A sleeve for a personnel air lock, the inside diameter of which is 9'10", is provided at centerline eleva-tion 30'6". Thicknesses of the primary components are as follows: .

Component Thickness (inches)

Barrel 3 1/2 1 3/8 Spherical Fl ange 5 3/8 Air . lock mounting 1 1/2 sleeve .

The equipment hatch cover is fitted with two seals that enclose a space which can be pressurized to 52.0 psig. The flange of the cover plate is at- The tached to the hatch barrel with 32 swing bolts,1-3/8 inch in diameter.

barrel, which is also the sleeve for the equipment hatch, is embedded in the shell of the concrete containment. The equipment h;tch' cover can be lifted to clear the opening. '.-

Inserted into.the' mounting sleeve through the equipment hatch cover is a personnel air lock consisting of two air lock doors, two air lock bulkheads, and the air lock barrel. Signi ficant dimensions of the air lock are as follows: -

l '

Parameter Dimension Inside Diameter of Barrel 9'6" '

Barrel Thickness 1/2" Door Opening 6'8" x 3'6" ,

Door Thickness ,3/4" .

Bulkhead Thickness 1-1/8" Each door is locked by a set of six latch pin assemblies, and is designed to withstand the design pressure from inside the containment. To resist the tes.t pressure, each door is fitted with a set of cast clamps. The doors are hinged and both swing into the containment. Each door is fitted with two seals that 1

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The doors are mechanistically interlocked so that only one door can be opened at a time. The capability exists for bypassing this interlock to equalize the pressure by use of special tools. The doors may be operated mechanically. .

Pers nnel Air 1.ock The personnel air lock (Fig. 3.3) consists of the air lock doors (2) and the lock barrel. The barrel, which is also the sleeve for the personnel air i lock, is imbedded in the shell of the concrete contaiment. The centerline of l .the oarrel is located at elevation 29'6" and an azimuth of 315*. Significant dimensions are as follows:

Parameter Dimensions Clear Opening 7'0" 0.D. of Flange on Door 7' 9 1/8" Bar.ci Thickness 5/8" Cover Thickness 5/8

The air lock barrel has a door on each end, each of which is designed to withstand the design pressure from inside the contairment. The doors are .

hinged and swing away from the air lock barrel. Each door is fitted with two seals that are located such that the area between doors can be pressurized to 52.0 psig. The locki ng device for the doors is a rotating, third ring, breach-type mechanism. These doors are also mechanically interlocked so that only one door can be opened at a time. The capability exists for bypassing ,

i this interlock and relieving the internal pressure by use of special tools.

The doors may be operated mechanically. .

N I Piping Penetrations There are two types of piping penetrations: moderate energy and high ene rgy. Moderate energy piping penetrations are trsed for process pipes in

- which both the pressure is less than or equal to 275 psi, and the temperature of the process fluid is less than or equal to 200*F. High energy piping pene-trations are used for that piping in which the pressure or temperature exceeds these values.

High energy piping penetrations (Fig. 3.4) consist of a section of pro-cess pipe with an integrally-forged fluid head, a containment penetration sleeve and, where a pipe whip restraint is not provided, a penetration sliding

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support inside the containment. The sliding support provides shear restraint while permitting relative motion between the pipe and the support. The annu-lar space between the process, pipe and the sleeve is completely filled with fiberglass thermal insulation. The pipe and the fluid head, are classified as ASME III Safety Class 2 (NC), whereas the sleeve is classified as part of the concrete contairinent, ASME III (CC). The sliding support inside the contain- /

ment is classifiea as an ASME Safety Class 2 component support (NF).

Moderate energy piping penetrations (Fig. 3.5) consist of one or more process pipes, the contairnent penetration sleeve, and a flat circular end-plate. The pipe.is classified as ASME III Safety Class 2 (NC). The sleeve is classified as ASME III Div. 2 (CC). The end-plate is classified as Class MC.

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Table 3.2 gives a list of the containment piping penetrations'. Included in this table is the penetration size. All of these piping penetrations are in the lower portion of the structure. '

4 -

Electrical and Instrumentation penetrations

!' Electrical pen:trations (Fig. 3.6) consist of a stainless steel header ,

i plate with an attach'ed teminal box, electric 1 modules which are clamped to

' the header plate, and a carbon steel weld rir.3 whica is welded to the header plate and to the sleeve. The metallic pressure resisting parts, the sleeve, i

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stainless steel header plate and carbon steel weld ring were designed as ASME III Safety Class MC components (NE); that portion of the sleeve which is backed by concrete was designed as part of the concrete containment, ASME III (CC).

' Double silicone and Hypalon 0-rings provide a seal with a cavity for

leakage monitoring between the header plate and the modules. The header plate is provided with a hole on the outside of the contalment to allow for pressurization of the penetration assembly for leakage monitoring.

There are a total of 64 electrical penetrations out of which 14 are spare and 8 are unused. All of these electrical penetrations are below the grade. '

j Instrumentation penetrations are of two types -- electrical and fluid.

i The electrical type is similar in construction to the other electrical pene-

- trations. The fluid penetrations are similar in construction to the' moderate energy piping penetrations.

! Fuel Transfer Tube Assembly The fuel transfer tube asse,51y consists of the fuel transfer tube, the i penetration sleeve, the fixed sat.'e on the reactor side, and the sliding sad-die in the fuel storage buildin<;. The fuel transfer tube centerline is at elevation (-)9'4-1/4" and it has approximately 20" inner diameter. Thq fuel transfer tube wall penetration sleeve, which is embedded in the concrete, has an inside diameter of a5out 25". ,

Ventilation penetratioris There are two' types of ventilation penetrations -- the contairnen* air

. purge penetrations (HVAC-1 and HVAC-2) and the containment on-line penetra-I tions (X-16 and X-18h The contairveent air purge penetrations (Fig. 3.7) each consist of a pipe sleeve (a rolled and welded pipe section, 36" outer diameter l by 1/2" wall thickness)' which is flanged at each and with 36" weld neck l

flanges and, attached to these flanges, the inner and outer isolation valves.

Together with the pipe, these valves fom a part of the contairment pressure boundary. The valves are 36" diameter butterfly valves with fail-safe pheu-i matic operators. The weld between the pipe and' the containment liner is .

! equipped with a leak chase for pressure testing. j The containment on-line purge penetrations each consist of a pipe sleeve l

(a rolled and welded pipe section, 8" o.d. by 1/2" wall thickness). A short section of pipe with a nipple is welded ta the sleeve on the outside of the.

contaiment, and a 3/4" valve and test connection is attached to it. The 9

I

Table 3.2 Containment Liner Penetrations Penetration Penetration Numbers Service Size Main steam line X-1 to X-4 30" X-5 to X-8 Main feedwater 18" X-9, X-10 RHR pump suction 12" X-11 to X-13 RHR to safety injrction 8" X-14 to X-15 Containment building spray 8" X-16, X-18 Containment on-line purge 8" X-17 Hydrogenated vent header 2" X-20 to X-23 CCW supply and return 12" X-24 to X-27 Safety injec-icn 4" X-28 to X-31 CVCS to pump seal injr.ction 2" X-32, X-34 Drain line 3",2" X-33, X-37 CVCS 3" X-35, X-36, X-40 RCS test / sample control 1" or smaller

X-52, X-71, X-72 X-38 Combustible gas control 10" X-39 Spent fuel pool cooling -

2" - ,,

X-43, X-47, X-50 Instrumentation lines d)!<f(

X-57 X-60, X-61 From containment recirculation sump 16" X-62 Fuel transfer tube 20" X-63 to X-66 Steam generator blowdown 3" X-67 Service air 2" HVAC-1,2 Containment purge supply / exhaust lines 36" ,

X-19, X-41, X-42 X-44 to X-46, X-48 X-49, X-51, X-58 Spare /j)I' X-59, X-58 to X-70 1

l

. )

0

- , - - ._ c...- -,--..- . - . -. . _ . , , . - -

OUTBOARD .

CONTAINMENT WALL INBOARD

STAINLESS STEEL

. HEADER PLATE t LINER PLATE l

THERMAL -

Y INSULATING GASKET.

I ,

1-BOX MOUNTING RING CANISTER .

i - ;Lr

. g

. ll 1. -

l ll ?

l 11 CARBON STEEL WELD RING /

FIELD WELD SLEEVE

/

JUNCTION (TERMINAL) BOX . JUNCTION BOX Figure 3.6 Tyical electrical penetrat.fon. ,

.l j

t ISOLATION VALVES ENCLOSURE BUILDING CONTAINMENT WALL SLEEVE

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- A.:I*~25 ISOLATION VALVES

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Figure 3.7 Typical ventilation penetration.

e

121-ends of this resulting assembly are welded to 8" weld neck flanges which are through-bolted to the inner and outer isolation valves. These valves are 8" diameter butterfly valves having fail-safe pneumatic operators. The weld be-tween the pipe sleeve and the contaiment liner is equipped with a leak chase for pressure testing. These on-line purge penetrations are very similar to those for 36" lines shown earlier.

3.2.3 1.eakage Rate Calculation Under severe accident' conditions the pressure inside the contalment At these pressures, any

.quickly builds up in the range of 75 to 200 psi.

leakage through the contalment holes will essentially be choked.

leakage under choked flow condition is given as (Ref.10):

The 2(

k+1 f

W= k(k II A II) where W = discharge rate (kg/s),

A = leak area (m2 ),

P = absolute pressure (N/m2 ),

p = mixture density (kg/m3 ), and k = ratio of specific heat at constant pressure to that at constant volume. .

For air and water vapor mixture, k - 1.3. If the mixture density is expressed by perfect gas law .

p=h (2) where R = gas constant, and .

T = the absolute temperature, Then Eq.(1) becomes k+1 (3)

W= k(h)TT A The mass of mixture can be written as M = Yp (4) <

M=h ,

where V is the free mixture volume in the contaiment. Equations (3) and (4) can be combined to get the leakage rate, in terms of mass fraction, as 9

e

. i l

I I k+1 vTA (5) i f= k(kH) ,

Note that the leakage rate, when expressed in terms' of mass fraction, depends ~ j only on the leakage area. .

j For Seabrook-1, usirg V = 2.704x10 8 ft3 and T = 296 F, Eq.(5) gives Leakage Rate = 0.721 Ai n W/0 per hour (6) 2 where A i n is the leakage area in in . Al ternately.

Leakage Rate = 17.3 Ai n w/o per day. (7)

The essentially intact e ntaiment leakage of 0.2 W/o per day, b thus, corresponds to an eq ent leakage area of 0.012 in2 2 (or, an equiva-lent hole of 1/8-in diameter). A leakage area of 4 to 10 in would correspond to the leakage rate of 2.9 to 7.2 w/o per hour. In other words, it will take '

2 about 14. hours to leak the entire content to the envirornent through a 10-in hole.

3.2.4 Containment Failure Model .

3.2.4.1 Leak-Before-Failure During accident sequences involving core damage, the containment struc-ture will be exposed to pressures and temperatures bey'ond those used in the design basis accident (DBA). Response of the contalment building to these severe conditions is evaluated in SSPSA by employing, for the first ti.me, a leak-before-failure model . In this model allowanceThis is made for continuous mode of containment leakage from the contaiment to the surroundings.

failure is temed local failure. The containment leakage can occur at many locations and discontinuities such as mechanical and electrical penetrations, personnel lock, equipment hatch, fuel transfer tube, welds, and in between the -

liner and concrete. Depending upon the size of leakage area and the accident sequence, local failures may gradually relieve pressure,, thereby gross con-taiment failure may be averted.

The leak-before-failure model is a realistic one. The extent of leakage and the health consequences must, however, be carefully studied. . In order to explain this issue, it is observed that traditionally probabilistic risk as-sessment is made by using what is temed a threshold model. In the threshold model, the containment is considered intact until the internal loading equals or exceeds a pressure threshold (which may also be temperature dependent), at which it is deemed to have suffered a failure (gross). If the internal load-ing is below this threshold value, the containment is considered intact and '

hence the risk is quite low. In the leak-before-fail'ure model, the release of activity, which is considerably small compared with that for the gross failAre mode, must be considered in health consequences. However, such leakages can potentially prevent the internal pressure from approaching the threshold value and thus a catastrophic or gross failure may be avoided.

9 e

. < .~23-

~

3.2.4.2 Classification of Failure The SSPSA report has clas,1fted containment failures in three categories:

. Containment Failure Category A. Inh.ludescontainmentfailuresthat -

develop a small leak that is substantially larger than the leak ac-ceptable from an intact containment, but not large enough to arrest ,

.the pressure rise in the contaiment. Category A failures thus cause i an early increase in the rate of leakage of radionuclides over the de- 1 sign basis leak rate but pressurization of the contaiment continues until either a category B or C containment failure occurs.

^

The intact containment is defined as the one in which leakage is lim- g g Cf:,

ited by the Technical Specification value. For Seabrook-1, this value is 0.2 w/o per day at the calculated peak accident pressure of approx- /

a *+udy has used 0.1 volume perf- g}

imately 47 psig. Note that the 5 cent per day for this leakage, a1though sne most recent amendment dated August 1984, the ;...jprior wcited both 0.1 volume per-cent and 0.1 w/o per day. The 10CFR50, Appendix J mandates the allow-able leakage to be quoted as w/o per day. The higher value noted here is based on Amendment 53, August 1984.*

. Containment Failure Category B. Includes failure modes that develop a -

large enougn leak area so that the pressure in the contaiment no l

longer increases. The time during which a substantial fraction of the i radionuclide source tenn is released is longer than approximataly 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Category B failures include self-regulating failure modes where the leak area is initially small but increases with pressure so i

that it becomes sufficient to terminate the pressure rise before a category C contaiment failure occurs. ,

The definition of " substantial" fraction is unclear.

k

{ . Contaiment Failure Category C. Includes those contaiment failure modes tnat develop a large leak area. A large fraction of the total radionuclide source tenn is released over a period of Wsm 1/ 1 L ..

hour. All gross failure modes are included in category C.

Mathematically, these three failure categories can be expressed in terms of leakage areas as follows: ,

Type A l pg_ 08A A 9 AA ANP AB'< Type B (8) l j AP Type C

y* _where. eM*

AC ,

ADBA = leakage area corresponding to th,e technical l

< f[' jf specification limit for containment leakage, j

Tnere appears to be substantial update / changes in thgAngineered Safety -)

l Features flow diagram, includingarrangements of motofopera ed valves and (.

bypass lines', which may substantially change the)requency o events. BNL,.

i however/is not rediewing-this -

part of SSkjSA.

9

y s (t j Anp = leakage area not large enough to arrest

, ] s1 g pressurizat,f on, and i h Ap 4 leakage area sufficient to release 100 w/o .

h*s \

N O in one hour.

The leakage area required to release a subst al fraction of the radio-9, Jnuclidd source term in approximately an hour an > can be computed using Eq.

(6). Assuming one-hundred percent turnover as a substantial fraction 2 in one hour Eq. (6) gives the required leakage area to be equal to 138 2 in or about pA p.

1ft. h Therefore, any containment leak area in excess of I ft will be -

(j' kA "\

fined as a gross _.contalment- failure -(Catecory. C)- This estimate of the ea '

area is factor two too high from t'he value stated in m g The leakage area required to arrest containment pressurization is in the range of 4 to 10 square inches, the lower value being more representative of

wet sequences and the upper value is representative of dry secuences. A leak

' area of about 6 square inches will result in the release of about 100 w/o of ac-ivity in a dag (see Eq. 7). Tne upper bound leak area for Type A failure is taken as 4 in . This corresponds to release of the radioactive source term (100% turnover in about 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The Category B leak area is, thus, in the range of 4 in} to 1 ft2 . Figure 3.8 is a pictorial representation of these, leakage categories.

' 3.2.5 Containment Pressure Capacity ,

3.2.5.1 Concrete Containment t

l Tne Seabrook PSA has examined failure modes for the containment structure i

itself, the steel liner, all penetrations, equipment and personnel lock hatch-

, es and the secondary cantainment. The containment structure includes the cylindrical wall, the hemispherical dome, the base slab and the base slab and containment wall junction. The most critical membrane tension was found to occur in the cylinder in the hoop direction. The median pressure which causes yield of both the liner steel and the reinforcing bars was found to be approx-l imately 157 psi, with a coefficient of variation of 0.084 The ultimate hoop load in cylinder is 216 psig. The contairunent wall is, tnus, assumed to fail

. at this pressure. At pressures beyend this, very large irreversible deforma-l tions occur which will cause cracks in the reinforced concrete but the lossThe of l

integrity of the pressure boundary may not occur until the liner tears.

i compiled radial deformations of the containment wall are shown in Figure 3.9.

i Note that the radial strain at the expected failure pressure of 216 psi is 4.7% ( Ar/r).

The hemispherical dome was calculattd to yield at a slightly higher pres-sure (163 psig). The failure pressure is predicted at 223 psig. .

I The median pressure for flexural failure of the base slab is 400 psig, 1 with a logarithmic standard deviation of 0.25. However, the shear mode of '

I failure is more restrictive. For this mode, the median failure pressure is estimated in SSpSA as 323 psig, with a logarithmic standard deviation of O.23. Although the uncertainty for failure of the base slab is large, the

probability of failure is small because the median capacities are high. Thus',

f ailure of tne base slab is not considered to be a critical failure mode and a

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an estimation of leak areas was, therefore, not considered for 'this mode of

< failure.

~

Secondary stresses in the cylindrical portion of the contalment occur at discontinuity such as at the base slab containment wall junction, at the springline, and where the amount of reinforcing changes. The flexural yield at the base of the cylinder occurs at 175 psi. At higher pressures, a plastic '

hinge .foms with considerable cracking of the concrete. These cracks, how-ever, are small enough so as not to threaten the integrity of the liner. The loss of integrity of the liner is not expected until a median pressure of 408 psi is reached. Thus, the failure of the base slab and containment wall junc-tion is not limiting.

i In summary, the containment wall is expected to undergo significant de-formation (=4.7% Ar/r) prior to its failure at 216 psig.

At this pressure.

Type C-(i .e., gross) f ailure occurs.

3. 2. 5.2 1.i na r_

The elongation capacity of the steel liner is computed by neglecting the-f riction forces between the liner and the concrete. The possibility that the liner stresses and strains could be different between two different pairs of tees was, however, considered. The SSPSA computed an elongation of 8.1 per -

. cent under uniaxial conditions, or an elongation of 4.7 percent under plane, strain conditions can be achieved without fracture. This would ensure integ-rity of the liner until fracture of the reinforcing bars. Additionally, the leakage of the containment at penetrations is considered likely before hoop failure of the liner occurs. '

3.2.5.3 Penetrations At all major penetrations, the containment wall is thickened and addi-tional reinforcement is provided to resist stress concent.ations. None of the l

meridional or hoop reinforcing bars are terminated at penetrations Instead, they are continued around the penetrations, thus ensuring that excess hoop and

": meridional capacity is available. Table 3.2 lists all piping penetrations.

,t '

As the contaiment pressure increases beyond its yield value (157 psi),

large radial defomations begin to occur. This induces stresses in the pipes by relative displacements between the containment wall and the pipe whip re-straints. Therefore, the most critical penetrations Also, are the areas where the stronger and . stiffer pipes pipe is supported close to the penetration.

develop higher forces at the penetrations for a given relative displacement.

The SSPSA study selected the following penetrations for investigation as being among the lines most likely to fail:

Penetration X-23 12" schedule 40 carbon steel (also X-20 to X-22 by -

similarity) /

Penetration X-26 4" schedule 160, stainless steel (alsoX-24,X-25,X-27) .

9 o

l

Penetration X-71 la - multiple pipe penetration (also X-72 and possibly*

others)

Penetration X-8 18" main feedwater schedule 100, (also X-5 to X-7) carbon steel Fuel Transfer Tube Convoluted Bellows The probability of failure at these penetrations was computed by (a) establishing a pressure-displacement relation, (b) estimating the failure probability as a function cf radial displacement and then (c) combining the two. The radial displacements for the containment wall were shown earlier (Fig. 3.9). The vertical displacement due to meridional strains is small (less than 3 in.hes) and hence its impact on the penetrations was ignored.

Since most of these. penetrations are in the lower part of the containment, the radial displacements experienced by them due to plastic deformation of containment would also be small. .

The multiple penetration (X-71 and X-72) would not fail even for the most unfavorable forces which these pipes could sustain. For penetrations X-23 and X-26, the most likely location for failure is at the partial penetration fil- -

let welds which join the pipe to the end plate. When failure of this weld oc-curs, the pipe remains in the hole provided in the end plate. The gap between the pipe and the end plate is likely to remain small unless the pipe wall buckles. Exact gap size is hard to compute. The SSPSA appears to usa a uni-form gap size of 0.04 in., and 0.10 in, as median and upper estimates, respec-tively. The corresponding leak areas for X-23 (as well as X-20 to' X-22) and X-26 (as well as X-24, X-25, and X-27) penetrations are shown -

in Table 3.3.,k s The dian failure ressure for X-23 penetration, t n n s e is higher than the hoop failure ressu e (216 psig) of he containment wall. These leak areas, therefore, are not expected to devel-op.

Penetration X-26 is expected to ' fail at a median pressure of 166 psig. ,-

The combined leak area for all safety injection penetrations 2 is obtained by independently adding individual median leak area of 0.5 in ,

Penetrations X-71 and X-72 are not likely to contribute to the averall leak area, as stated earlier. .

The main feedwater lines (penetrations X-5 to X-8) are 18-in. diameter, Schedule 100 pipes. The failure mode of most concern is failure of the flued At a median pressure Mhe d due to axial loads in the pipe at the penetration.

ofU/180 psig,2each one of these Since all fourpenetrations of these canisfail likely to result in independent of ae.ach leak area of 50 in each.

other, the total leak area is 200 in2. Although the failure of a single such penetration can be cortsidered as Type B failure, is all four main feedwater penetrations were to fail simultaneously the resulting leakage will be of Type f<

C.

The fuel transfer tube is fixed to an elevated floor inside the contain-ment. As the pressure in the containment increases, the containment wall-moves outwards and thereby exerts pressure on the bellows. The most pertinent

'. . 'o Table 3.3 Leak Area Estimates for Mechanical Penetratio'ns

+

Median Median Line Size Penetration Leak Area Failure Pressure in2 p3tg .

6.0 >216 12" X-20 to X-23 CCW Supply and Return 2.0 166 4" X-24 to X-27 Safety Injection X-71 and X-72 Negligible je 1" Sample / Control X-5 to X-8

~

200 180 18" di Main Feedwater Fuel Transfer Tube 3 172 --

X-16, X-18 See Text 8"

On-line Purge HVAC-1,2 See Text . 36" Containment Purge s

t 4

O

-m

bellows from the viewpoint of. containment leakage is the one inside the con-tainment (EP-2).- Three potential failure modes, in their order of decreasing

' probability of failure, considered are (a) failure due to overall buckling of the bellows, (b) failure due to local buckling within the convolute, and (c).

failure due to meridional z bending strains. The SSPSA hasThis estimated median is a Type A leak area of about 3 in at a pressure of about 172 psig.

failure.. ,

There are two sets of containment penetrations which are cpen to the l

contaiment atmosphere on the inside. The on-line penetrations (X-16 .and X-18) are the 8-inch purge suction and discharge lines and contaiment purge suction and discharge lines (HVAC-1 and 2) are the 36-inch lines. Each one of these four lines has two containment isolation valves, one inside and one ,

outside the containment. All eight valves are pneumatically operated butter-3 i

fly valies. At elevated temperatures, the seal material (usually ethylene

propylene) on thes2 valves may deteriorate and lose its sealing function.

Any deposition of radioactive aerosols could further deteriorate the sealant materi al . Consideri.,g sealant degradation due to temperature alone, ethylene

! propylene seal life (Ref.10) is 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , 40 mts, or 20 mts if exposed to 400, 500 or 600 F, respectively.

In the event of the failure of the sealant menei to,an crack leak.

leak into the space path may develop cnd containment atmosphere may l between the two isolation valves. Since the isola on valves are closed from the containment isolation signal system, the leakage of containment atmosphere '

l to the environment can occur only if the sealant of the outer containment iso-lation butterfly valve also fails. The time duration elapsed before this happens can be significantly long (of the order of hours). The SSPSA has es-timated it to be long compared to the containment failure by other causes.

The SSPSA study, therefore, has disregarded this release path.

The available leakage area due to sealant degradation has been estimated (Ref.10) by assuming an equivalent clearance of 1/16 inch between valve disc and body for ' low' and 1/8 inch for 'high' estimates. This gives a total '

leakage area of 17 in z as low value and 34 in as high value. As noted ear z

lier, the outer butterfly valves must also experience high temperatures prior E<

to a through release path. This leak area is of Category B. The SSPSA study

. has argued that such a leak path is not likely to result prior to a gross 4 - cor.taiment failure (Category C).

Electrical penetrations can fail primarily due to overheating of the pot-ting compound. Tte S$PSA study has concluded that the failure of electrical pe7etrations is not expected to make a significant contribution to containment ]

failure for any accident sequence. This conclusion, appears justified for the l wet case, but, for the dry case, it is based on their estimate of slow over- gf heating of the potting compound. A careful Such thermal conduction calcula' tion a calculation, similar to the <

i should be made to check this assessment. )

problem of vent / purge line butterfly isolation valve failure, is beyondj the scope of this work and hence it was not done. ~

The equipment hatch and personnel lock penetrations can fail either due to pressure loading or degradation of the sealant material (generally silt -

cone). The structural failure, prior to containment failure, appears unlike-

) ly. The sealant material can degrade at high temperatures typical of a

severe accident. According to the 0-Ring Handbook (see Ref.10), s'ilicone can survive for twenty hours when exposed to 500 F temperature. Furthemore, the personnel air lock is a double door system so even if the sealant around one door were to become ineffective, substantial time delay would be required to make the second sealant also ineffective. It, thus, appears that the equip ~

i ment hatch and personnel lock penetrations do not contribute significantly to Type 8 failure.

3.2.5.4 Containment Failure Probability

- The calculation of the probability of containment failure as a function of the pressure is quite involved. The method used and results reported in the SSPSA study seem reasonable except for the impact of all four main feed-water lines failure. The SSPSA has categorized the failure of X-8 (one of the four main feedwater lines) penetration as Type S rince anticipated leak area is 50 in 2. It appears to us that when one such penetration fails, the remain-ing three will also fall at nearly the2 same pressure of 180 psig (195 psia).

Any depressurization due to a 50-in hole is not likely to be fast enough to reduce the containment pressure substantially prior to the failure of the l three remaining penetrations. Assuming that all four main feedwater lines fail at 180 psig, an equivalent leak area of 200 in z will result. This fail- !

ure, therefore, should be classified as Type C. The impact of this change on

l the containment failure probability numbers will be to reduce the rate for Type B with a corresponding increase in Type C. zThe total failure rate is not likely to change. Estimated containmanc failure fractions are compared with

the SSPSA results in Fig. 3.10.

3.2.5.5 Containment Enclosure

.. g The containment enclosure building is designed to withstand 3.5 psipres-sure difference between the enclosure and the environment. Du ri ng normal The operation, the internal pressure .is about -0.25 iriches of water gauge.

SSPSA study has calculated its pressure capacity to range from more than 1 psid to 10 psid. In view of relatively strong primary containuent, the role of the secondary containment is important primarily for Type O failures of the primary containment. In the event of Type C failure, the secor.dary enclosure building might not play any significant role as far as the source term calcu-

! lation is concerned.

3.3 Definition of Plant Damage States and Containment Resconse Classes _ -

The grouping of accident sequences into plant damage states proceeds from the premise that the broad spectrum of many plant failure scenarios can be discretized into a manageable number of representative categories for which a ;

single assessment of core and contaiment response will represent the response l

of all the individual scenarios in that category. .

The plant damage states classify events in accordance to the following three parameters: /

Initiating Events

1. !

. "A" - Large loss of Coolant Accident j "S" - .Small Loss of Coolant Accident "T" - Transient i 1 j '

i I

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,6.: :-

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e I l 1 500 ,2 2,0 : ,5;i . g ip o 14 0 lito : .

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Figure 3.10 Estimated containment failure fraction. j .. .

, l

. . - *s . ...

a

2. Timing of Core Melt and Conditions at Vessel Failure .

"E" - No RWST Injection to RCS "L"" With RWST Injection to RCS o~ -

- N Emergency Feedwater g[//

( "FW" - Emergency Feedwater .

3. Availability of Containment Systems ,

i "C" - Long-Tenn Containment Spray Cooling "4" - Long-Term Spray Recirculation, No Cooling "1" - Isolation Failure er Bypass I Figure 3.11 gives the definition of the plant damage states and their re-spective frequencies listed in Table 3.4 as used in the SSPSA risk model.

These damage states are categcrized in a matrix of eight physical conditions in the containment (numerals (1) to (8)) and six combinations of containment safety function availability (letters A to F) for a total of 48 potential plant damage states'. A ninth damage state type has been defined for accident sequences involving steam generator tube ruptures. Figure 3.11 indicates that only 39 plant damage states can be identified as credible sequences.

~

From the viewpoint of containment response, many of the plant damage states can be grouped into contalment classes. The classes defined in Table

! 3.5 are differentiated primarily according to spray availability. The fre .

quency of each containment class is the sum of the frequencies of the plant l

states included therein. .

Annual plant state frequencies calculated by the applicants for both in-ternal and external events were reviewed by the Lawrence Livermore National Laboratory 6 and were found acceptable. Table 3.6 presents, the calculated contaiment class frequency estimates for internal events, fires, floods 'and truck crashes; moderate and severe seisaic events, in order to comprehensively assess the risk from seismic events, it is necessary to make separate consequence calculations for those accidents which are initiated by earthquakes severe enough to impair evacuation. For this '

l purpose, the seismic frequency estimates are divided into two categories in l Table 3.6. The seismic events with instrument peak ground acceleration below 0.59 can be binned with internal events, fires, ficods and- truck crashes.

Seismic uation, andevents'with acceleration must be treated greater separately in thethan 0.50g areanalysis.

consequence judged to impair evac- M' These contaiment response classes ( Q ant damage states) are the starting point for the contairment event nreejnalysis and they define the link or interfaces with the plant analysis _ y 3.4 Contaiment Event Tree and Accident Phenomenology An important step towards the development of the contaiment matrix in-volves the quantification of branch point probabilities in the containment event tree. These probabilities depend heavily on the analyses of degraded--

and core melt phenomenology and the contaiment building response described in Appendix H of the SSPSA.5

. . 34

,,. .. . ... .. .. o ,..,

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any m e m e e e

'a - 8 @ 8 @ O 8 8

-~ - 395; gg:g e sig e e y e e e e e e w. -

Q& .Q vitb M %d:: ham ~8 e @ @ O O O 8 M h 8 @@R O O m ;r -

~~

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Plant State Rep resents 2A AEC 4A TEC SEC 2C/6C AE4 4C/8C TE4 10 AE 20/60 AL 30/70 SE,TE/TEFW 4D/80 SL,TL 2E/6E AECI 4E/8E TECI IF V 2F/6F AEI 3F/7F SEI 4F/EF SLI .

1 l

l l

Figure 3.11 . Definitions of the plant damage states used in SPSS.

i

~

- 35-Table 3.4 Frequencies of Occurrence of the Plant Damage States Frequency Frequency

Plant Damage (events per Plant Damage (events per State reactoryear) State reactor year) 10 .3.03(-7) 6A 3.41(-7)

IF 1.89(-6) 6C 3.57(-10) 4 IFA 6.10(-11) 60 2.49(-7)

IFP 8.52(-7) 6E 5.30(-14) 2A 1.85(-6) 6F 2.08(-16) 2C 1.91(-9) 6FA 1.11(-11) 20 2.53(-7) 6FP 1.34(-12) 2E 1.40(-13) 70 7.06(-5) 2F 1.06(-13) 7F 3.55(-8) 2FA 3.10(-11) 7FP 1.09(-5)

+

2FP 1.58(-10) 8A 4.50(-5) 30 1.94(-5) 8C 4.29(-8) 3F 5.00(-7) 80 5.51(-5) '

3FP 6.21(-6) 8E 5.02(-11) 4A 1.28(-5) 8F 1.02(-10) 4C 1.65(-7) 8FP 1.95(-7) 40 2.79(-6) 9A 7.51(-10) 4E 2.24(-11) 9C 3.62(-13) 4F 2.25(-13) 90 9.09(-9) 4FP 1.18(-7) ,

TOTAL 2.30(-4)

NOTE: Exponential notation is indicated in abbreviated form; i .e. , 3.03(-7) = 3.03 x 10-7, .

1 9

w a

S Q

t

- - - - - -,,e-- - - ,,e,--w. ,-a,--,-- - ~ - , - - + - - - , ----,.,--,,,-,.,-,---,-,,,---.--,-m.+-----,,--,----,--,-----,n,

l Table 3.5 Containment Response Class Definitions

~

Class Plant State Represents 1 1D AE 2 2A/6A,4A/8A AEC, TEC, SEC 3 2C/6C,4C/8C AE4, TE4, SE4 .

t 4 30/70 SE, TE, TEFW 5 20/6D,4D/80 AL, SL, TL 6 -

IF, 2F, 3F, 4F, 6F, V 7F, 8F 7 2E/6E, 4E/8E AECI, TECI 8 IFP, 3FP/7FP Small leaks w/o RWST ,

9 2FP/6FP,4FP/8FP Small leaks w/ RWST 10 IFA, 2FA/6FA Aircraft crashes.

11 9A V2 (SGTR) 12 9C V2 (SGTR) 13 90 V2(SGTR) t O

e e

~

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I .

Table 3.6 Containment Class Hean Frequenciest Frequency (per reactor year)

Containment Internal, Fires, Internal Response Class Floods and Truck Seismic <0.5g Seismic >0.5g Total Seismic and i Crashes External I 1 1.08E-7 - 1.95E-7 1.95E-7 3.03E-7 2 5.70E-5 1.54E-6 1.24E-6 2.78E-6 6.0E-5 3 1.80E,7 1.91E-8

1.91E-8 1.99E-7 4 8.60E-5 1.85E-6 2.27E-6 4.12E-6 9.0E-5 5 5.50E-5 1.10E-6 1.76E-6 2.86E-6 5.8E-5 , .

6 1.80E-6 1.66E-7 3.93E-7 5.59E-7 2.4E-6 7 '

4 8

5.29E-6 1.25E-5 1.79E-5 1.79E-5 9

1.12E-7 2.40E-7 3.52E-7 3.52E-7 10 * -

L y

11

1 12

13 -

l 1 Reference [5] Tables 5.1-3 and 9.2-9.

~

j

Indicates frequencies less than.10-8 yr l.

i ..

l .

1 x_

. . 238-The SSPSA contalment event tree uses the twelve top events identified in Table 3.7 as major phenomenological phases which could occur with respect to the formation and lo,ca,t o , core debris. These processes are grouped into four phases folloyi' 9 cident initiation (1) phenomena occurring while the core is still n 1 ; (2) phenomena occurring while the core is located-be'ow tne lower grid p ate but is still in the reactor vessel; (3 ena occurring with the core debris located in the reactor cavity a d onthe n-tainment floor; and (4) the phenomena involving long-tenn coolin co n-tainment and/or basemat penetration.

3.5 Containment Matrix (C-Matrix)

Tne twelve top events in the Seabrook containment event tree are summar-ized in Table 3.7. A negative response at any of the five nodes (4, 8,10, 11, and 12) in the containment event tree results in the failure of the con-taiment buticing by a variety of failure modes. Each of these failure modes

['

results in a particular radiological release category. For those paths that do not have a negative response at any of the five nodes, the path will even-tually result in no failure of the containment. The conteinment event tree thus links the plant damage states to a range of possible contalment failure modes via the various paths through the tree. For a given tree, each path ends in a conditional probability (CP) of occurrence, and these cps should sum to unity. The quantification of an event tree is the process by which all the -

paths are combined to give the conditional probabilities of the various release categories. In SSPSA, fourteen release categories are used for the quantification as summarized in Table 3.8. Note that two of these release categories (namely, SS and S3) correspond to intact / isolated contai nment.

Fission product release for this category would, therefore, be via normal leakage paths in the containment (ar.d rinclosure) building, which can be if= ~ 3 ferent-depending on availability of the enclosure building v'entilatio fil-

't atjon system.

Table 3.9 sets forth' a simplified containment matrix (C-matrix) for the Seabrook plant using the contaiment response class definitions discussed in Section 3.3, and the release category definitions given in Table 3.8. In arriving at the C-matrix of Table 3.9 all of the very low probability values were disregarded. This is shown7 to be insignificant to the risk estimate.

The present assessment of containment response for Seabrook plant is not based upon independent confinnatory calculations of accident progression and contalment response. Instead the knowledge gained from review of similiar risk studies for other l

.3. , pressurized water reactors with large dry containments is used to guide this assessment. .

The mode and timing of contaiment failure cannot be calculated with a great degree of accuracy. Judgements must be made about the nature of the dominant phenomera and about the magnitude of several important paramet6rs.

Furthemore, the codes and methods used for these calculations are approximate <

deld do not model all of the detailed phenomena. Fortunately, risk measured in personal exposure is not sensitive to minor variations in failure mode' and.

timing. It is important, however, te prcperly cha.acterize the major attri-butes of failure mechanisms; (1) whether the failure is early or late, (2) whether it is by overpressurization, bypass, or basemat melt-through and, (3)-

whether or not radionuclide removal systems are effective.

7 Table 3.7 Accident Phase and Top Events for the Seab, rook Containment Event Tree

~

Accident Phase Top Event j Initiator 1 Plant State D'ebr i in Vessel 2 Debris Cooled in Place 3 No H2 Burn i 4 Containment Intact Debris in Reactor Cavity 5 Debris Dispersed from Cavity

6 Debris Cooled

. 7 No H2 Burn 8 Containment Intact Long-Term Behavior 9 No Late Burn 10 Containment Shell Intact 11 Basemat Intact Failure Mode 12 Benign Containment Failure (Small Leak) 6 6

i .

l I w

}

4 i

.~,- -.,-. . ~ . , . - - -,_.n,_,._,_.,-,- , . , . . - - . . . . , . . ~ , , - . ~ - - . . . , , _ , , . . , . - - - , ,,,_,_,,_,,--_,,,,.,,-,.,,....,--,,,nn, ~, - . ,n.,

.. 40 Table 3.8 Release Categories Employed in the Seabrook Station Risk Model Release Category Release * '

Group Category Definition 55 Containment intact / isolated with enclosure air handling filtration working.

Containment Intact / Isolated .

55 Same as 55 but with enclosure a r handling filtration not working.

S2 Early containment leakage with late over-pressurization failure and containment building sprays working.

U Same as 52, but with containment building spray not working.

4 Tf7 Same as W, but with an additional vaporiza-

, tion compcnent of the source term.

S3 Late overpressurization failure of the con-

! Long-Term tainment with no early leakage and contain-Containment ment building sprays working.

! Failure -

U Same as 53, but with containment building sprays not working.

T37 Same as U, but with an additional vaporiza-tion component of the source term.

54 Basemat penetration failure, sprays operating

, S4V Containment basemat penetration failure with

containment building sprays not working and additional vaporization component of the source term.

S6 Containment bypass or isolation failure with containment building sprays working.

T57 Same as 56, but with containment building -

sprays not working and an additional vapori- '.

Early zation component of the source term. 4 Containment .

Failure / Bypass 51 Early containment failure due to steam explo-sion or hydrogen burt),with containment ,

building sprays working, f

~~

H Same as S1, but with containment building sprays not working. l

S denotes applicability to Seabrook Station; number corresponds with contain-ment failure mode; bar denotes nonfunctioning of containment building sprays; ,

, and V denotes achievement of sustained elevated core debris temperatures and !

associated vaporization release.

4

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

< Table 3.9 Simplified Centainment Matrix for Seabrook 4

Release Category Class 51 52 53 55 56 W U 3W 3Ti M 3Ti 55v-d 1 . 0.60 0.40 0.01 0.99 2

3 1.0 4

0.89 0.11 5 1.0

6 1.0 i 7 1.0 '

i 8 1.0 9 1.0 i

10 1.0 11 1.0 12 - I. O 11 1.0

.__. \

1 0

, -y- .

9 4

.e' me O

e 9

e > .

f- ,

The assessment of the containment response and failure mechanisms is based on the general understanding of that accident phenomenology and the con-l tainment design characteristics discussed earlier. The phenomena of interest 4 may be summarized as follows: 7 .

i i Early Failure (51, TT) which fo result fr~n a steam explosion or an early hy-drogen burn is believed to unlikely. Altnough explosions in the reactor probable, the resulting mechanical energy would veste1 1'o wer plenua are .

1 be limited by the fraction of the core which could participate in a sing"le,7ex- we plosion and by the efficiency of the p.'ocess. In recent PRA reviews, have assigned a conditional probability of 10-4 to steam explosion induced i contaiment failure. This probability lead: to the conclusion that steam ex-j plosions would have a negligible effect on risk, and consequently, the appli- i j cants 5x10-* value is not included in the simplified C-matrix.

The conditional probability for an early containment failure due to ex-4 l

ternal e nts (i.e.; aircraft crashes) is assigned 1 in the SSPSA as shown in i Table . This simply indicates that an aircraft crash ir.to the contalment is assumed " J. toD~ faill the containment structure with certainty.

1 .7 3 (

Early containment failure could also conceivably result from direct heat-

! ing due to a rapid dispersal of the core debris throughout containment in the .

fonn of aerosols. The dispersal could only be caused by the high primary sys-l tem pressures that may exist at vessel failure for a number of transient se-i quences (recent calculations 11 indicate that there exists a propensity for

! establishment of natural convection pattern inside the reactor vessel and the hot leg; which can cause rapid heatup of the RCS boundaries possibly leading to failure and depressurization prior to bottom head melt through, thus elim-inating, high pressure ejection sequences). The aerosols coul,d rapidly pres-

'surize containment by direct heat exchange and concomitant chemical reac-

, tions. Scoping calculations performed by the Containment Loads Working Group l

(CLWG) showed that a very severe challenge to the containment integrity could i

result provided 25 percent of the core mass were converted to aerosols.12 i

However, no consensus could be reached among the CLWG analysts as to the cred-

) ibility of this parameter value, and this failure mode is still speculative. -

l Furthennore, the configuration of the Seabrook lower cavity would tend to re-i duce the dispersal of core debris beyond the reactor cavity boundaries.

I For the reasons outlined above as well as the high contai_nnent failure i

pressure for Seabrook, it is concluded that early overpressure failure has a very low likelihood.

l h Early Containment Leakage (S2, 37, "T27) without gross failure of containment

< building is expected to occur for nonisolated steam generator tube rupture

, , event with containment sprays available (S2), for large break LOCA sequences

)

with RWST injection in the absence of sprays (T2), and for dry cavity i sequences with a vaporization release.(T2V). .-

l There seems to be a basic inconsistency in assigning plant damage stittes

to this failure mode as defined in the C-matrix. Specifically, large breal(~

j LOCA sequences with, RWST injection in the absence of contaivaent sprays are i expected to lead tc. an D failure mode with 100Y, probability (see 37 below);

i 1

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

while they are also assigned to TE with 1007. probability. This ca'n be correct only if the initiator and the sequences are indeed different, but at this time i we cannot resolve the inconsisi:ency. .

Similarly, the significance of contaiment functions on steam' generator- /J

' tube rupture sequences is not at all obvious.

Late Overpressuriza*. ion Failure (53, 57 TR) can occur due to steam produc-ticn in a wet cavity or noncondensable gas production as a result of core-con-crete interaction for a dry cavity situation. For sequences in which early and intemediate failure is not expected to occur, and for which contaiment sprays are inoperable, failure is expected to be a certainty.

The conditional probabilit. - late overpressurization failure with a

' vaporization release (dry cavityJ .s down to be 0.60. This results from tt.e i relative competition between the late overpressure failure and the basemat 4 penetration (3Ti) for accident sequences without the containment sprays.

l The failure time for the late ovirpressurization failure mode is much longer than previously calculated for other large dry co ntai me nt .1,3 ,i.

This is as a result of the very high failure pressure for the Seabrook con-J taiment. As a consequence of this high contaiment failure pressure (median ,

pressure of 211 for wet and 187 psia for dry

sequences) it is difficult to challenge the contaiment integrity by any conceivable event.

J Hydr: gen deflagration early in the accident sequence or later after vessel failure when steam condensation occurring as a result of reactivation of sprays (due to regaining of ac power), or other natural heat sink mecha-nisms, which can produce a deinerted atmosphere is not expected to challenge the contalment integrity.

The impact of changes in the contaiment failure distribution discussed in 3.2.5.4 is not significant for late failures. .

Basemat Penetration Failure (54, T47) can only result' in the absence of con-taiment neat removal system (sprays) for a dry cavity. A 26-inch high curb surrounds the reactor cavity that prevents the entry of water into the cavity unless the full RWST has been injected. The conditional probability of the basemat melt though is always less than the late overpressurization failure, particularly for Seabrook with the natural bed rock fomation directly under the basemat foundation. Therefore, the basemat penetration failure probabil-ities are con,servatively assigned, i

No Failure (SS, TT) would result for all sequences with full spray operation.

Tne raoiological releases are thus limited to the design basis leakage with essentially negligible off-site consequences. .

Contalment isolation Failure (56, 5TO is represented by an 8-inch diameter

" purge line. Ine accident sequences where the containment is either/not 4

For dry sequences, only primary system water inventory is available in the

.. contaiment. In this case, the contaiment atmosphere becomes superheated '

and, at f ailure, the temperature can exceed 700'F.

, ~ . - - - , - - , , . . - - - , , _ . - - . - - , - - - , . _ , . - _ - . , - . - , - , _ , , , , . , ._--,.--.-.--.,_n. . . - , , - - . _ - , , - - . .

isolated or bypassed (Event V) are assign 2d a conditional probabil.ity of unity to this release category.

An 1.nterfacing systems LO'CA (V sequence) results from valve disc rupture or disc failing open for series check valves that nomally separate the high pressure system. This event results in a LOCA in which the reactor coolant bypasses the containment and results in a loss-of-coolant outside the cor.tain-ment. Furthemore, the concurrent assumed loss of RHR and coolant make-up capabil~ity leads to severe core damage. In the SPSS, three possible inter-fae ~ag systens LOCA sequer.ces have been found and discussed. These are

1. Disc rupture of the check valve in the cold-leg injection lines of the RHR. .

2. Disc rupture of the two series motor-operated valves in the nomal RHR hot-leg suction.

3. Disc rupture of the motor-operated valve equipped with a steam mount-ed limit switch and " disc failing open while indicated closed" in the other motor-operged valve in the nomal RHR hot-leg suction.

For the V-sequence, the core melts early with a low RCS pressure and a dry reactor cavity at vessel melt-through. The containment sump remains dry ,

anc recirculation is not possible.

The core and con;aiment phenomenology used to arrive at the split frac-tions for the contairsnent event tree and thus the' C-matrix are in general agreement with the other previously NRC reviewed studies 1 ,3, for PWRs with large dry contairments. Furthemore, the claimed unusually high strength of the Seabrook contairnent reduces the impact of sensitivity caused by uncer-tainties in the severe accident prog ression. However, should the claimed stre gth of the contairment be reduced to levels comparable to some of the other large dry containments, the impact of ~ uncertainties may become signifi-cantly more pronounced, as discussed in our review of the MPSS-3.7 3.6 Release Category Frecuencies Based on the contalment class frequencies in Table 3.6 and the contain-ment f ailure matrix of Table 3.9, the release frequencies were computed and are summarized in Table 3.10. /f Table 3.10 indicates that on1 light f the release categories dominate

, the total release f requency.

Tables 3.11 and 3.12 set forth the contribution to : ore melt frequency from the various contaltnent response classes and release categories, respec-tively. It is seen that containment classes 2, 4, and 5 dominate the core melt frequency whil l the release categories SS (containment intact), U and ,

D Y doninate the sou c tem frequency.

I 1 .

Table 3.10 Frequency of Dominant Release Categories (yr-1)

Internal, Fires, -

Floods and Truck . Internal and Category Crashes Seismic <0.5g Seismic >0.59 External ,

S3 7.50E-7 3.45E-8 2.69E-7 1.05E-6 55 5.64E-5 1.52E-6 1.23E-6 5.92E-5 TE - ,

1.12E-7 2.40E-7 3.52E ,7 TJ 5.50E-5 1.10E-6 1.76E.6 5.79E-5 Tfi 5.29E-6 1.25E-5 1.78E-5 Tf7 7.66E-5 1.65E-6 2.14E-6 8.04E-5 T4Y 9.50E-6 2.04E-7 3.27E-7 1.0E-5 T67 1.80E-6 1.66E-7 3.93E-7 2.36E-6 O

e 4

e

( . .

i

~

Table 3.11 Centributlen of Containment Response Classes to the Total Core Melt Frequency Internal, Fires, Internal Containment Floods and Track and Class Crashes Seismic <0.5g Seismic >0.5g Total Seismic External 1 - - - - <0.01 2 0.25 <0.01 <0.01 O.01 0.26 3 - - - - <0.01 4 0.37 0.01 0.01 0.02 0.39 .

5 0.24 - 0.01 0.01 0.25 6 0.01 - - - 0.01

)

8

0.025 0.055 0.08 0.08

~

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l I

e ,

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4 i

Table 3.12 Release Category Frequency as a Fraction of Core Melt Frequency ;

i Release Internal. Fires, Internal Category Floods and Truck and ;

!l Crashes Seismic <0.5g Seismic >0.59 Total Seismic External

. 53 <0.01 <0.01 <0.01 <0.01 <0.01 i '

55 ~

~ ' 0.25

<0.01 <0.01 0.01 O.26

.l ~

I $2 * <0.01 <0.01 <0.01 <0.01 i 37 0.24 ,

<0.01 <0.01 0.01 0.25 ,

TfiI -

0.03 0.05 0.08 0.08 4 TW 0.33 0.01 0.01 0.02 0.35'

$4V 0.04 <0.01 <0.01 <0.01 0.04 b.

i j 3W 0.01 <0.01 <0.01 <0.01 0.01

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4 ACCIDENT SOURCE TERMS 4 ,

In this chapter the approach utilized in the SSPSA to determine the '

i fraction of fission products originally in the core which can leak to the out -

side enviroment will be outlined. The fission product source to the environ-l ment as calculated by this approach for each release categtry will also be .

l 1 discussed. ,

j 4.1 Assessment of Severe Accident Source Terms As in the Reactor Safety udy (RSS)l3 the CORRAL-!! code was used in the k SSPSA for determining fission (product leakage to the enviroment. This code i

' takes input from the thermal-hydraulic analysis carried out for the contain.

f ment atmosphere. In addition, it needs the time-dependent emission of fission products. The fission products were assumed to be released in distinct phases ;

as suggested in the RSS, namely, the Gap, Melt, and Vaporization phases. The i

[

time dependence of these phases is detemined by the timing of core heatup, primary system failure, and core / concrete interaction. The methods used in '

4 the SSPSA differ from the RSS methods in the following ways:

1) The treatment of iodine was changed and iodine was treated as cesium iodide. This was accomplished by merely using the same fraction of

core inventory released for both the cesium group and the iodine group, ,

2) Leakage releases are represented by a multi-puff model.

3) An uncertainty analysis was carried out in which it was attempted to

I 1

account for shortcomings in the RSS methods.

i i

j In general, the not result of the SSPSA analysis was to reduce the fractional release of particulate fission products. This will be discussed in scre de- (

tait later. In all, fourteen releases were detennined ranging from contain- [

. ment bypass sequence to the no-fall sequence as shown in Table 3.8.

l These release categories were evaluated by considering the contaiment '

1 failure mode, the availability of the spray system, and whether or not the j cavity was wet or dry. Table 4.1 shows the point-estimate releases as deter- t i

mined by the methods outlined above. Containment failure mode Si corresponds l 1

to a gross failure of the contaitunent, resulting from a steam explosion, early l pressure spikes, or early hydrogen bu ns. Failure mode 52. represents a loss of contalment function early in the ccident sequence. This loss of function ,

takes the form of an increase in the leak rate to 40% per day where it stays l l

i until the contalment fails due to overpressurization. Failure mode 53 repre-sents a late overpressurization failure of the containment driven by dalcay l

heat or late hydrogen burns. Failure mode 54 represents a basemat melt-

- through, 55 represents no containment failure and the leak rate is limited to .

!, the contaiment design basis leak rate. Finally, failure mode 56 represents l

sequences where the containment is failed or bypassed as part of the initiat-J ing event.

i The second parameter considered in defining the source term is the avail

ability of sprays. This is determined by the plant damage states. Those j

l

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

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T l I2 99999 M M99 MM M M *. ** 3 y

massa =,,~,, . ss: == sms as M-.M M~ s: .

a ,,,, , ,,,i ~~ -

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89.*,9.9 a 9999 *M 9999 . M* M9~M s 4

.F]# .; . .=,. .z' -* x - a s - a s-- 1r

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release categories with operating sprayg systems are designated 51 to 56, while those with spray systems,not operating are designated 3T to TT.

The third and final parameter considered in dif ferentiating between source terms distinguishes between wet and dry cavities. In the case of dry' cavities a vapo,rization release due to core / concrete interactions will occur, while for wet cavities the core debris is assumed to be quenched or the water .

in the' cavity will scrub the vaporization release thus effectively reducing the release to zero. The release categories which include a vaporization re-lease include a "1" in their designation as shown in Table 3.8.

From the point of view of risk it was found 5 that 57T, NT, IN, and T67 were dominant either for acute or latent health effects. In view of this re-suit these four categories will be considered in more detail.

Release categories 37 and TW have late everpressurization failurf modes, with no spray systems operating and differ only in the omission or inclusion of a vasorization release, respectively. The containment at Seabrook is cal-culate: to fail at a median pressure of 211 psia for vet sequences and 187 psia for dry sequences. At this pressure a gross failure is expected result- From ing in a puff release of approximately 0.5 hr release duration.

Table 4.1 it is seen that the IT and IN sequences fail at 27.2 hrs and 91.5 -

hrs, respectively. These failure times are severalThe hours later than was cal-primary reason for the culated for Indian Point, Zion, and Millstone-3.

later failure in this case is due to the super'or strength of the containment-

~

structure. Table 4.2 compares the !3, 5W release parameters with similar parameters for the other three reactors mentioned above. Note that a fair comp arison should set (0!+1) equal to (Cs-Rb), since iodine was treated as Cs!, It is seen that I, Cs, and Ba groups for 5T are approximately half the other releases, while the Te, Ru, and La groups are low by apsroximately an order of magnitude. This difference is due to the latter fatlure time, allowing more time for settling and the absence of a vaporization release.

which dominates the release of Te, Ru, and La. A similar comparison for the

- TI7 release indicates a unifom reduction of approximately an order of magni-tude for all species. The reduction is entirely due to the late failure time for this sequence. ,

Another important consideration is the increased rate of release due to an increase in the leak area prior to attaining gross failure conditions.

This :an also impact the radienuclide transport mechanisms inside the contain-ment due to changes in the containment themel hydraulic conditions.

Release category 377 is associated with early contalment failure in which the contalment function is compromised by increasing the leakage area in such a way that the leak rate increases from 0.1% per day to 40% per day.

This release rate is not enough to prevent an ultimate overpressurization '

failure. This release is modeled as a the multi-puff

release. The first puff corresponds to the release up to time when vaporization st'a rts (melt + gas). The secon pu cludes the period of vaporization release -and the third puff is equi to an overpressurization failure at the time of catastrophic contalme failure. In this model the duration of the melt.

' \ ,

f(J .

9

Table 4.2 Late Overpressurization Failure Comparison 7

Millstone-3 Zion / Indian!" Indian 3 Seabrook 5 Point Study Point

~

37- 3W M-7 TML8' 2RW l.

Xe 9.0(-1) 1.0 9 (-1) 9.6(-1) 1.0

)

10+1 1.2(-1) 2.4(-2) 1.5(-1) 1.05(-1) 9.3(-2)

Cs-Rb 1.2(-1) 2.4(-2) 3.0(-1) 3.4(-1) 2.6(-1) l

) Te-Sb 2.2(-2) 3.0(-2) 3.0(-1) 3.8(-1) 4.4(-1) j Ba-Sr 1.5(-2) 2.6(-3) 3.0(-2) 3.7(-2) 2.5(-2) !

Ru 4.4(-3) 2.3(-3) 2.0(-2) 2.9(-2) 2.9(-2) ,

1 La 4.4(-4) 3.9(-4) 4.0(-3) 4.9(-3) 1.0(-2)

T (release) 27.2 81.5 20 (hrs)

T (duration) 0.50 0.50 0.50 - 0.50 *

(hrs) -

i

' Energy 300E7 300E7 540E6 150E6 j (8tu/hr) . :

)

i

. I i

l l

release is s n to be 3.5 haurs, vaporization release 7.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .and the re- '

maini ng ase 78.0 hours0 days 0 hours 0 weeks 0 months . It is not clear that the melt release in this case is 5.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , however, it does not seem to be unreasonable. A 7 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months which duration r the vaporization release is not consistent with the RSS,N only allows 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for this phase. Finally, it is not clear how the 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months . 6 for tr.e last phase was deterv.ined. The release duration for a single puff, which is he sum of the above three phases leads to a release time of 88.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months whicr seems extraordinarily long. Our recommendation would be to reduce 2Q these times to be more consistent with RSS methods (see Table 4.7).

The total rele.:se of fission products from the. sequences can be compared to the M-4 release determined for the M111 stone-3 study. This comparison is made in Table 4.3. It is seen that, once adjustments are made for the dif-farent ways in which iodine is treated, the IIT release ,is approximately half the M-4 re12ase. Without .the benefit of a However, calculation, it is difficult to a possible reason for judge whether the differences are reasonable.

this reduction is the credit taken for the enclosure building surrounding the actual containment building. This feature is unique to the Seabrook contain- .

ment structure.

Release ca*egory T5V has binned into it an isolation failure corresp nd-i ng to an 8" diameter breach in containment and the interfacing LOCA (V-sequence). This sequence is also represented by a multi-puff release. In

this case as in the previous case, the total release time is long compared to acceptable limits of the RS$ 13 consequence mo:al. Our recommendation would be

/fj to reduce these times to more reasonable values (see Table 4.7).

The release fraction can be compared (Table 4.3) to the M-4 release from M M111stgne-3, PWR-2 for the RSS and the V-sequence from the RSSMAP study for Surry. 5 Excep'; for the iodine group, it is seen that the release fractions are comparable. If the iodine group were set equal to the cesium group value, it is seen that th value for T67 would be the lowest release fraction.

4.2 Source Tenn Uncertainty Analysis In this section we will briefly describe the uncertainty analysis carried '

out for the four dominant accident sequences and, where possible compare the fission product leakage to the environment to more mechanistic determina-tions . There are two contributors to the uncertainty in release characteriza-tion. First, uncertainty in time parameters which are influenced by:

1) Prediction of key event times, and

2) The mix of accident sequences binned into a release category.

Second, uncertainties in release fractions, which are influenced by:

1) Analysis methods and data, and - .

2) Uncertalrties in timing of key events.

8 6 e

~

Table 4.3 Comparison of Releases for Failure to Isolate Centainment and the By-Pasi 5equence ,

7 Seabrook5 Millstone-3 RSS13* RSSMAP15 IUFT 1RPT M-4 PWR-2 V-Sequence 4

Xe l.0 9.7(-1) 9.0(-1) 1.0 1.0 0!+1 3.1(-1) 4.3(-1) 2.0(-1) 7.0(-1) 4.8(-1)

Cs-Rb 3.1(-1) 4.3(-1) 6.0(-1) 5.0(-1) 7.9(-1)

Te-Sb 3.2(-1) 4.0(-1) 5.0(-1) 3.0(-1) 4.4(-1)

Ba-Sr *4(-2)

. 4.8(-2) 7.0(-2) 6.0(-2) 9.0 ('-2 )

Ru 2.5(-2) 3.3(-2) 5.0(-2) 2.0(-2) 4.0(-2)

La 4.2(-3) 5.3(-3) 7.0(-3) 4.0(-3) 6.0(-3)

T (release) 2.2 2.2 0.20 2.5 2.5 (hrs)

T(duration) 88.7 14 2.0 1.0 1.0 (hrs) .

Energy (8tu/hr) 140E6 4E6 70E6 20E6 0.5E6

The same as M1A release category in M111 stone-3.7

.Y '

..- s. 2,O j .

-7' e

e e

0

$, The above principles were used to determine source term multipliers which l

roduct leakage to the environment. A probabil- ,

would give a rangewith of fission j ity ,is associated each p" source term, and for later overpressurization f ailure modes (D, 'UV, and 577) the following discrete probability distribu .

tion is used, i.e.,

. Subcategory probability .

i

,! U-a .02 i U-b .08 U-c .30 ,.

37-d .60 This indicates, for example that there is an 8% confidence level that U-b correctly defines the source term for the IT release category.

The results of this analysis for the overpressurization failure modes is: ,

4 I

! Particulate P.elease Factor (multiplier) ,

J Probability IT M 3N -

02 .22 .63 .17

.071 .22 .07 ,

l .08 *

.024 065 .02 i

.30 .007

. .60 .0071 .021 i 4

From this table it is seen that for the most likely release, i.e., "d", the

- reduction factors of the source tenn are substantial.

]

The first two releases can be compared to releases published in 8MI-2104 l Volume V (Surry) for the TML8'-c and AS-c sequences. These two sequences t l

j correspond to late containnent failures and are both binned into 53 and M [

sequences. A comparison of these sequences 1,s shown on Table 4.4. From this l j{ table it is evident that for the volatile species Xe, Cs, and I, the release l i categories U and IW bracket er exceed'the mechanistic estimates carried 'out in 8MI-2104 for both TM:8' and AS sequences. However, for the less volatile j species Te, Ba, Ru, and La, the release of the AB sequence is the only one [

bracketed or superseded by the IT and 5W releases. The release fraction

{

determined for the TML8' sequence is higher than all the IT and M relea'ses. l This discrepancy is primarily. due to the comparatively early failure time. It ;

is felt that agglomeration and settling would reduce the source for the TML8' sequence to values close to those reported for ITand M. No comparative sequence for 3 W was analyzed in SMI-2104.

In the case of the 3U release category a different probability distribu-tion was used. This change reflects the break location, which initiates the l

i

) e I l '

f l

Table 4.4 Comparison of AB-c and TR B'-c (OMI-2104) to 3 W and 37

~

Release Fractions -

Release Probability ~ Release Category Time (hrs) Xe Cs I Te ta Ru La Tff-a .02 28 1.0 1.5(-2) 1.5(-2) 1.9(-2) 1.6(-3) 1.5(-3) 2.5(-4) 3Ti-b .0a 36 9.0(-1) 5.3(-3) 5.3(-3) 6.6(-3) 5.7(-4) 5.1(-4) a.6(-5)

~

3 W-c .30 54 8.0(-1) 1.6(-3) 1.6(-3) 2.0(-3) 1.7(-4) 1.5(-4) 2.5(-5) .

TTi-d .60 89 7.0(-1) 5.0(-4) 5.0(-4) 6.3(-4) 5.5(-5) 4.8(-5) 8.2(-6) ,

5e

\

1 ~

l 3T-a .02 22 1.0 2.6(-2) 2.6(-2) 4.9(-3) 3.3(-3) 9.7(-4) 9.7(-5) 3T-b ,.08 28 ' 9.0(-1) 8.5(-3) 8.5(-3) 1.6(-3) 1.1(-3) 3.1(-4) 3.1(-5) 3T-c .30 34 8.0(-1) 2.9(-3) 2.9(-3) 5.3(-4) 3.6(-4) 1.1(-4) 1.1(-5) 3T-d .60 53 7.0(-1) 8'.5(-4) 8.5(-4) 1.6(-4) 1.1(-4) 3.1(-5) 3.1(-6)

TE R*-c -

12 1.0 2.S(-3)- 6.0(-4) 8.5(-2) 1.7(-2) 2.4(-5) 4.3(-4)

A8-c .

24 1.0 4.8(-5) 4.7(-5) 4.0(-5) 4.9(-5) 2.4(-7) 3.6(-5) t .

I * * ;

4 . .

l V-sequence. This break could be either in the hot-leg (b release subcategory) or the cold-leg (c release subcategory). This sequence is modeled as multi-puff release and each puff is treated separately. In this comparison only the l

4 sum of the release will be considered, since no adequate method of analyzing a

l l

multi-puff release is readily available. Table 4.5 shows a comparison be ' l tween the totals of tre various SET releases and two V-sequence releases com-puted for Surry and published in BM!-210c. One of t te V-sequences is " dry,"

1

[

l

] implying no water in the path of the rel:sse and the other is " wet," implying ,

i that the release passes through 3 feet of water before entering the atmo- l

! sphere. From this comparison it ,an be seen that all the releases, except Cs l i

for the " dry" V-sequence, are bracketed by the 35V releases.

l i

I

) 4.3 Recommended Source Terms u

The severe accident source tems used in the Seabrook Probabilistic Safe. !

j

ty Study reviewed in the previous sections, are aimed at the multi-puff con-sequenc

model present in the CRACIT computer code. In order to make these

{ source tems useful to the NRC staff for evaluation with the CRAC code, total J

Furthemore, the suggested i l releases must be used as sucmarized in Table 4.6.

source tems of Table 4.6 together with their release category characteristics l l given in Table 4.7 have been adjusted to more closely represent our assessment jy ,

I of,the severe accidents based upon the RSS methodology.

// t It must also he noted that the suggested source tem for the Steam Gener- l i ator Tube Rupture ($GTR) sequence is assumed to be one-tenth of the source '

tem for the event V (ITV). This is believed to be a conservative estimate and can be used in the absence of a more specific mechanistic calculation. l The suggested source tems can be used to estimate the number of health i

! and economic effects (consequences) in,the population surrounding the Seabrook

) Station due to radioactive atmospheric releases as a result of core melt acci- ;

's ;

j dents. ,

The resulting consequences together with the frequency of radiological f l

1 releases will enable the establishment of the severe accident risk at the Sea-

) brook site considering the double-reactor unit effect.

!' . i i

4 f

! i i t i I t

' r l

L___ !

Table 4.5 Comparison of W (sum) to V-sequence (Surry)

Release Fractions Release Probability Cate9ery ,Ie Cs I Te la Ru La W -a .02 .97 4.3(-1) 4.3(-1) 4.06(-1) 4.2(-2) 3.32(-2) 5.3(-3)

W -b 45 .97 2.95(-1) 2.95(-1) 1.36(-1) 3.53(-2) 1.52(-2) 2.0(-3)

W -c 45 .97 1.295(-1) 1.295(-1) 3.2(-2) 1.593(-2) 5.2(-3) 5.3(-4) e 10 W -d .08 - . 97 5.2(-2) 5.2(-2) 1.3(-2) 6.6(-3) 2.0(-3) 2.2(-4)

V (dry) 1.0 * *

5.52(-1) 1.99(-1) 1.2(-1) l 1.0 * *

V -

1.04(-1) 3.84(-2) 2.5(-2)

(submer9ed) .

Indtvidually not reported.

e

?

t . c

?

i

'. ~

Table 4.6 BNL-Su99ested Scurce Terms Release La Category Xe 0! I-2* Cf Te Sa Ru 0.94 - 0.023 0.023 0.24 0.0033 0.41 9.8E-5 .

51 . !

52 0.89 - 2.1E-5' 2.1E-5 4.4E-6 2.9E-6 8.8E-7 8.8E-8 53 0.90 7E-3 1.E-7 1.E-7 1.9E-8 '1.3E-3 3.8E-9 3.8E-10 e

l- i' (in. 0.0091 - 3.5E-8 3.5E-8 6.1E-9 4.0E-9 1.2E-9 1.2E-10 55 0.90 - 3.6E-3 3.6E-3 6.7E-4 4.4E-4 1.3E-4 1.3E-5 0.94 - 0.75 0.75 0.39 0.093 0.46 2.8E-3 ,

10I 0.90 - 0.31 0.31 0.057 0.038 0.011 1.1E-3 ICI .

0.31 0.31 0.32 0.034 0.025 4.2E-3 Il5I 1.0 -

lli 0.90 - 0.12 0.12 0.022 0.015 4.4E-3 4.4E-4 377 1.0 - 0.024 0.024 0.030 2.6E-3 2.3E-3 3.9E-4 1

1P0T 1.0 - 0.058 0.058 0.072 6.2E-3 5.4E-3 9.1E-4 l

55 0.014 7E-4 5.2E-7 5.2E-7 9.5E-8 6.3E-8 1.9E-8 1.9E-9 0.97 0.43 0.43 0.40 0.048 0.033 5.3E-3 137 337-d 0.90 - 0.043 0.043 0.040 4.8E-3' 3.3E-3 5.3E-4 l

Elemental todine not used, all todine treated as Cst.

l

**3E7 d release is 1/10th of the II7 values. , ,

a 6

i 4

+

e

.. .., j 59- ,

.~ ,

Table 4.7 SNL-Suggested Release Characteristics for Seabrook - ,

Release Release Release Release Warning +

Release Time Duration Energy Het ht Time

()

Category (hr)

(hr) (Stu/hr) (hr) .

0.35

)

51 1.9 0.5 140E6 10 ,

52 2.6 1.0 0.5E6 10 1.05 !

~

64.0 0.5 250E6 10 43 53 , ;

10 0.35 >

o '- l 55 1.9 10 n/a 56 4.5 4 0.5E6 10 ,

0.5 0 +

TT 1.4 0.5 520E6 10 0.30 TT 27 10 10E6 10 26 .

10 3G IN 35 10 25E6 ,

0.5 250E6 10 26 >

TT 27 ,

76 TN 81 0.5 450E6 10 0 49 TW 50 0.5 250E6 ,

h 4.3 24 10E6 10 0.30 TW 2.5 1.0 0.5E6 10 1.0

'~

TW d 2.5 -

1.0 0.5E6 10 1.0

Warning time is defined as the time efter core melt starts to the time of radiological release.

-3' e

f e

l e

a L

. ?

Io .

. .'60-en n

5.

SUMMARY

AND CONCLU510NS

' - The purpose of this report is to describe tne technical review of the Seabrook Station Probabilistic Safety Assessment aM to present an assessment of containment perfomance, and radiological source tern estimates for severe -

core melt accidents.

The contairment response to severe accidents is judged to be an important factor in mitigating the severe accident risk. There is negligible probabil-

' ity of prompt contaiment failure or failure to isolate. Failure during the F'. first few hours after core melt is also unlikel Most core melt accidents would be effectively mitigated by contalment spr y operation, i

, Our assessment of the contalment failure aracteristics indicate that,

' there is indeed a tendency to fail contaiment through a realistic benign mode compared with the traditional gross failures.

The point-estimate release fractions used in the 55PSA are comparable in jcases magnitude to those used in the R$5. In those where comparisons can-be made to the more mechanistic source term study carried out by the Accident

- Source Tem Program Of fice (ASTP0) and repor ed in BM!.2104 it was found that

b. the 55P5A releases were either higher than o for the most part similar to the

j '. recent release fractions. It was also fou that the energy of release was somewhat higher in the 55PSA than for other misting studies.

}

i

/9'

-MM 1 4 WN 4 .u

!, m. "yJ

}

A .Lu , WA H - t ,y,

'\

j 'd C.emfartson to h/ ASH /420 feinte Ty~ n .

74l me h v. .f cin - _

9 4cm,

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rk

&&=f .s 4 I I .1.en M e d si gg . 73 '

j i ouurm

%L 9 a e i

i 3. Lb4 c.,dah 8 O N .

1-i I

1

6. REFERENCES

1. " Zion Probabilistic Safety Study," Commonwealth Edison Company (September -

1981).

2. " Limerick Probabilistic Safety Study," Philadelphia Electric Co.

(September 1982).

3. " Indian Point Probabilistic Safety Study," Power Authority of the State of New York and Consolidated Edison Company (March 1982).

4. " Millstone Unit 3 Probabilistic Safety Study," Northeast Utilities (August 1983).

5. B. J. Ga rrick, et al . , "Seab rook Station Probabilistic Safety Assessment," Pickard, Lowe and Garrick, Inc., PLG-0300 (December 1983).

6. A. A. Garcia, et al., "A Review of the Seabrook Station Probabilistic Safety Assessment," Lawrence Livennore National Laboratory Report (Dec.

12,1984).

7. M. Khatib-Rahbar, et al., " Review and Evaluation of the Millstone Unit 3 .

Probabilistic Safety Study: Contai nment Failure Modes, Radiological

! Source Terms and Of f-Site Consequences ," NUREG/CR-4143 ( report to be published). ,

8. R. O. Wooten and H. Avei, " MARCH: Meltdown Accident Response Character-istics - Code Description and User's Manual," BMI-2064, NUREG/CR-1711 '

(1980).

9. J. F. Muis, et al . , "CORCON-Mod 1: An Improved Model for Molten ,

Core / Concrete Interactions," SAND 80-2415 (1981).

10. B. E. Miller, A. X. Agrawal, and R. E. Hall, "An Estimation of Pre-Exist-ing Containment Leakage Areas and Purge and Vent Valve Leakage Areas Re-sulting from Severe Accident Conditions," A-3741,11/15/84 (Draft report dated August 1984) transmitted via letter to V. Noonon, June 29, 1984

11. W. Lyon (organizer), "RCS Pressure Boundary Heating During Severe Acci-dents," USNRC Meeting, Bethesda, Maryland (May 14,1984),

12. " Estimates of Early Containment Loads From Core Melt Accidents," Con-tainment Loads Working Group, NUREG-1079 (Draft 1985).

13. " Reactor Safety Study," U.S. Nuclear Regulatory Commission, WASH-1400, '

NUREG-75/014 (October 1975).

~

14. . " Preliminary Assessment of Core Melt Accidents at the Zion and Indian .

' Point Nuclear Power Plants and Strategies for Mitigating Their Effects," l NUREG-0850, Vol.1 (November 1981).

15. G. S. Kolb,'et al., " Reactor Sa fety Study Methodology- Application'

! Program: Oconee #3 PWR P1 ant," NUREG/CR-1659/2 of 4.

, - , . - - - - - . , - - - . - ,- -. ,.. -,

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