ML20209E299

Summarizes Work & Findings to Date from Review of Probabilistic Safety Study in Response to J Rosenthal Request.No Deficiencies Found.Review Will Not Be Continued

ML20209E299
Person / Time
Site:	Seabrook
Issue date:	04/10/1985
From:	Hulman L Office of Nuclear Reactor Regulation
To:	Sheron B Office of Nuclear Reactor Regulation
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ML20209E304	List: ML20209E299 ML20211D986 ML20211E040 ML20211E144 ML20211E226 ML20211E331 ML20211E403 ML20211E458 ML20211E494 ML20211E574 ML20211E585 ML20211E591 ML20211E603 ML20211E610
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[un reegDo UNITED STATES

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%' O j/ APR 101985 YY 3, WI Docket No.: 53=469 MEMORANDUM FOR: m m Chief Reactor Systems Branch Division of Systems Integration FROM: L. G. Hulman, Chief Accident Evaluation Branch Division of Systems Integration

SUBJECT:

REVIEW 0F SEABROOK PROBABILISTIC SAFETY STUDY We understand that the subject review will not be continued, nor will we be expected to provide source term evaluations or consequence calculations in support of the Seabrook review. We are, however, summarizing our work and findings to date in the discussion provided below in response to the J. Rosenthal request for a brief summary. .

We previously gave informal comments to Warren Lyon on the preliminary containment response / source term review report submitted by Brookhaven National Laboratories.

The rest of our review was restricted to consequence calculation methodology. It appeared that the evacuation model in the Seabrook Probabilistic Safety Study was developed from a reasonable data base.

However, the complexity of the evacuation model and its interaction with the meteorological model in CRACIT warrants a more lengthy review than was justifiable (given the cessation of the review). We note, however, that impaired evacuation during summer weekends (because of tourists) that was an issue during licensing, was considered in the consequence analysis. No gross deficiencies in this or other aspects of the consequence calculations were found before the review was terminated.

. G. Hulman, Chief Accident Evaluation Branch l Division of Systems Integration uston J. Rosenthal W. Lyon 8 5LW DI CDb Vp ~

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) f1 f.g glt The corment numbers correspond to report sections identified with the same number:

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

i[ a. Containment design - reinforced, d

a 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, q (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.

l (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 result in a steady or decreasing containment pressure. This would I prevent failure of the other feedwater pipes. Even if all four feedwater

$ pipes were to fail, the corr.bined leak area would not be much larger than

% that for a single pipe failure. - The failure mode is self-regulating and the total leak area is determined by the pressurization source, i.e., the steaming rate or the gas generation rate.

(7) Page 30: Failureoftheouterpurgeisolation'valveandtheelectrica[

penetrations is very unlikely for the same reason. The heat losses into the concrete are too large to raise the temperature in the penetration j before mass transfer into the interspace stops. Steam condensation leads l to a noncondensable atmosphere in the interspace which is in pressure j equilibrium with the containment. From this point on, only conduction through the sleeve can transfer heat into the penetration.

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(8) Page 30: Electrical penetration integrity is based en:

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

.b . Calculated heat losses to the concrete. (See coment 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 and 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 gross 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 penetrations 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 S1 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 (P at VMT>181 psia)

Pr(SI) = Pr (P at VMT>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) Page 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 energy release which limits the plume rise to the inversion layer.

The latter limitation (2) was found more constraining. ,

(15) Page 52: The start of the second puff release coincides with the beginning of the vaporization release. The duration of the vaporization release is only 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> but the duration of the second puff is 7.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> -

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. ,

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O Page 52: We feel it would be-appropriate to point out that your (16) recommendation for reducing 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. We do not believe that the RSS 4

times are more reason *ble.

Page 56: To us the single puff release concept is'not a realistic one

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because we now know that all the risk significant releases at Seabrook extend over several hours. The purpose of an " equivalent" single puff release can only be to either assess the conservatism of a single puff release versus a multipuff release or to bring the analysis within the i

constraints of the CRAC code. In either case, this should be clearly I stated.

-(18) Pages 52 and 56: Table 4.7 lists BNL recomended release

' characteristics which are based on the MARCH / CORRAL calculated point

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estimates, without taking into consideration any of the advances since WASH-1400.

In the SSPSA source term uncertainty analysis, it is shown ,

that these release characteristics represent something like a 99%

confidence level for nQgexceedance. We think it would be appropriate to list in a separate table the single puff-equivalents of the SSPSA best

' estimate release categories. These would be T.T-d,'SlV-d, (T6V1-d +

56V2-d + 55V3-d) and (57V1-d + DV2-d + DV3-d). These could be listed simply for comparison without a recommendation, to give the future user i

a choice of using whatever he believes to be appropriate for his purpose.

(19) Pages iii and 60: We believe that the following two conclusions are

. equally significant:

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

~ .> o An analysis of the source term uncertainties has shown that the best

/ estimate source terms can be expected to be between one and two l

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 are,shown in the uncertainty analysis. Therefore, release category S2V-a on Tab _le 11.6-16 should bi used for this comparison. This release category (52V-a) reflects the l release timing calculated for this type of accident sequence using the i

WASH-1400 methodology.

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A REVIEW OF THE SEABROOK STATION PROBABILISTIC SAFETY ASSESSMENT: CONTAINMENT FAILURE MODES AND RADIOLOGICAL SOURCE TERMS t

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 O

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ABSTRACT 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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. ACKNOWLEDGMENT 1

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CONTENTS .

Page iii ABSTRACT................................................................

ACKNOWLEDGMENT.......................................................... vi iv LIST OF TABLES..........................................................

vii LIST 0F' FIGURES.........................................................

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

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1.1 Background....................................................

1.2 Ob j ec ti v e s a n d Sc o pe . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.3 Organization of the Report.................................... 1 2

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

2.1 As s e s sme n t o f Pl a n t De si gn . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 -

2.2 Compa ri s o n wi th Othe r Pl ant s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 7

3. ASSESSMENT OF CONTAINMENT PERFORMANCE.............................. .

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

3.2 Co n t a i nm e n t F a i l u r e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.2.1 Background............................................. 8 3.2.2 Design Description..................................... 8 3.2.3 Leakage Rate Calculation............................... 21 3.2.4 Cont ai nment Fa il ure Mode 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.2.4.1 Le a k-te fo re-Fa il ur e . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.2.4.2 Cl a ssi ficati on o f Fail ure. . . . . . . . . . . . . . . . . . . . . 23 3.2.5 Contai nment Pres sure Ca paci ty. . . . . . . . . . . . . . . . . . . . . . . . . . 24 3.2.5.1 Concrete Containment.......................... 24-3.2.5.2 Liner......................................... 27 3.2.5.3 ' Penetrations.................................. 27 3.2.5.4 Containment Fail ure Probabil ity. . . . . . . . . . . . . . . 31 3.2.5.5 Co n t a i nme n t Encl o s u r e . . . . . . . . . . . . . . . . . . . . . . . . . 31 3.3 Definition of Plant Damage States and Containment

- Re s p o n s e Cl a s s e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 3.4 Containment Event Tree and Accident Phenomenology.............

33 3.5. Containment Matrix (C-Matrix)................................. '38 1 3.6 Rel ea se Ca teg o ry Fr e quenci e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44

4. ACCIDENT SOURCE TERMS.............................................. 48 1

4.1 Assessment of Severe Accident Source Te rms . . . . . . . . . . . . . . . . . . . . 48-4.2 Source Te nn Unc e r tai nty Anal y si s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 4.3 Re c omme n d e d So u rc e Te rm s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,56

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SUMMARY

AND CONCLUSI0'NS............................................ 60 -

I 6.- REFERENCES......................................................... 61 ,

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• _ -vi-LIST OF TABLES Table Title Page 2.1 Compari son of Sel ected Design Cha racteri sti cs . . . .. .. . . . . . . . . . . . . . 5 3.1 Containment Ope ra ting and De si gn Pa ramet e rs. . . . . . . . . . . . . . . . . . . . . . 10 3.2 Con t ai nment Li n4 r Pene tra ti on s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . 18 3.3 Leak Area Estimates fc Mechanical Penetrations....... .... .... .. . . 29 3.4 Frequencies of Occurrence of the Plant Damage States............. 35 3.5 Contai nment Re s ponse Cl ass De fi ni tions . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 3.6 Co n ta i nment Cl a s s He an Fr e que nci 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 Model............................................................ 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 Contair. ment 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 Ov erpressuri zation Fa il ure Compa ri son.. . . .. . .'. . . . . . . . . . . . . . . 51 4.3 Comparison of Releases for Failure to Isolate Containment 53 andtheBy-Pjss5equence.........................................

I l 4.4 Comparison o AB- c and TMLB'- c (BMI-2104) to lET7 and T7. . . . . . . . . . - 55 4.5 Compa ri son of 1997 (s um) to V-sequence (Surry) . . . . . . . . . . . . . . . . . . . . 57 4.6 B NL- Su g g e s t e d So u r c e Te rm. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . SS j

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

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LIST OF FIGURES Figure Title Page 3.1 A schematic represen* ; tion of source term calculation............ 9 3.2 . Equipment hatch with personnel airlcck........................... 12 3.3 Personnelairlock................................................ .

14 3.4 Typi cal high energy pi pi ng penetrati on. . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.5 Typical moderate energy piping penetration....................... 16 3.6 Typi cal el ectri ;al pene tration. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.7 Typi c al ventil a ti on pene tration. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.8 A pictorial representation of laahage categories................. 25 3.9 Estimated radial displacement of containment wall................ 26 3.10 Estimated containment failure fractions.......................... 32 3.11 Definitions of the plant damace states used in SPSS..........;.... 34 l

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1. INTRODUCTION 1.1 Backcround

• 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 Brookhaven National Laboratory (BNL)

Regulatory Commission (NRC) for review.

under contract to the NRC, has been involved in reviewing core malt phenome-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 Yankee Atomic Electric Company in December 1983.5 1.2 Objective and Scooe The objective of this report is to provide a perspective on severe acci-dent propagation, containment 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 sequences were reviewed by '

the Lawrence Livermore National Laboratory (LLNL) as reported in an incomplete report [6] prepared for the Reliability and Risk Assessment Branch of NRC. -

7 , --In the pres report, principal contai nment design features are dis-cussed and compared w*th hose of Zion, Indian Point and Millstone-3 designs.

Those portions of th PSS Telated to severe accident phenomena, containment resptnse ar.d radial source terms are described and evaluated. Numerical 1 adjustments to the SPS jestimates are documented and justified. )$

1.3 Organization of t e Report At brief review o/f the Seabrook plant features important to severe acci -

dent analysis is prefented in Chapter 2 along with comparisons to Zion, Indian Point and Hillstone-3 plant designs. Chapter 3 contains the assessment of contai nment perfo rmpnce. Specifically, definition of ' containment response classes and plant d9 mage states, analytical .ehod containment failure model, containment event ree and accident phenomeno o y and the containment matrix are reviewed. Ch ter 4 addresses the accide t source terms together with justifications for adjustment where necessary. The results of this review are summarized in Chap' er 5. .

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-e-2, PLANT DESIGN AND FEATURES IMPORTANT TO SEVERE ACCIDENT ANALYSIS In this section, those pf ant design features that may be 'important- to an assessment of degraded and core melt scenarios and contaiment analysis are reviewed. These important features are then ccmpared with the Zion, Indian Point and Millstone-3 facilities to identify comonalities for . benchmark compa ri sons.

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 a re 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, Westir; house 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 dome 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 nment , protects -the liner over the foundation mat. The containment structure provides biological shielding for normal and accident conditions. The approximate dimensions of the containment are:

Inside diameter 140 ft.

Inside height 219 ft.

Vertical wall thickness 4 ft. 6 in, and 4 ft. 71/2 in.

Dome thickness 3 ft 6 1/8 in. '

Foundation mat thickness 10 ft.

Contaiment penetrations are provided in the lower portion of the structure,

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

Each contaiment enclosure (also known as secondary contair. ment) sur-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 leakage from the O contairment structure other than leakage associated with piping, electr'ical and access passage penetration and discharge to the filtration system of

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r co ntai ment. To accomplish this, the space betweert the contaiment enclosure and the containment structure, as well as the penetration and safeguards' pump.

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 ensure air tightness for the negative pressure, leakage through all , joints and penetrations has been minimized.

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-a-A containment spray system is utilized for post accident containnent heat renoval. The containment spray system is designed to spray water containing boron and sodium hydroxide into 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 containnent dome. After a prescribed )

amou,t of water is removed from the tank, the pump suction is transferred to the containment sump, and . cooling is continued by recirculating sump water through the spray heat exchangers and back through the spray headers.

The spray is actuated by a containment spray actuation signal which is gererated at a designated containment pressure. The system is completely re-dundant and is designed to withstand any single failure.

The containment isolation system establishes and/or maintains isolation of the containaant .from the outside environment in order to prevent the re-lease of fission products. Automatic trip isolation signals actuate the ap-propriate valves to a closed position whenever automatic safety injection oc-curs or high containment pressure is experienced. Low capacity thermal elec _

tric hydrogen recombiners are provided.

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

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 containmenc sump and the reac-tor Core.

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.

I The control building contains the bu'ilding services necessary for contin-uous occupancy of the control room complex by operating personnel during all operating conditions. These building services include: HVAC services, air i purification and iodine removal, fresh air intakes, fire protection, emergency breathing apparatus, communications and meteorological equipment, lighting, and housekeeping facilities.

Engineered Safety Feature (ESF) filter systems required to perform a j safety-related function following a design basis accident are discussed below:

l a. The containnent enclosure exhaust filter system for each unit col-l lects, filters and discharges any containment leakage. The system is not normally in operation, but in the event of an accident, . it' is placed in operation and keeps the contai nment enclosure and the l building volumes associated with the penetration tunnel and the ESF l equipment cubicles under negative pressure to ensure all leakage 'from.

the containnent structure is collected and filtered before discharge l

to the plant vent.

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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 with dampers and controls will maintain the building at a negative pressure.

The emergency feedwater system supplies demineralized water from the con-densate water storage tank to the four steam generators upon loss of normal 1 feedwater flow to remove heat from the reactor coolant system. Operation of the system will continue until the reactor coolant system pressure is reduced 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 soJrces to assure delivery of condensate under emergency conditions.

The two units of the facflity are interconnected to off-site power via three 345 kilovolt' lines of the transmission system for the New . England i

states. The normal preferred source 3f p wer 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 provided to isolate the generator fro:n 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-formers are not available, the redundant safety feature buses of each unit are I powered by two reserve auxiliary transformers. Upon loss of off-site power, l

each unit is supplied with adequate power by either of two fast-starting, diesel-engine generators. Either diesel-engina generator ar.d its associated <

safety feature bus is capable of providing acequate power for a safe shutdown under accident conditions with a concurrent loss of off-site, power. A con-stant supply of power to vital instruments and controls of each unit is assur-l ed through the redundant 125 volt direct current buses and their associated I battery banks, battery chargers and inverters.

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• 2.2 Comoarison with Other Plants .

! 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.

It is se'en that the contairinent characteristics are quite similar with 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 r.early iden-i tical between the four units.

The Seab rook 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 initjal gas consists largely of j

carbon dioxide, the quantity of which depends on the amount of calcium ca/ bon-(

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 I

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Table 2.1 Comparison of Selected' Oesign Characteristics Zion Indian Point Millstone Seabrook Design Parameters Unit 11 Unit 2 3 Unit 34 7 Unit 1,2 5 Reactor Power [HW(t)] 3,250 3,030 3,411 3,650 Containment Building: g Free Volume (ft 3) 2.73 x 10' '2.61 x 105 2.3 x 108 2.7 x 10' Destgr Pressure (psia) 62 62 59.7 6 .7 Initial Pressure (psia) 15 14.7 12.7/9.1 15.2 Initial Temperature (*F) 120 120 120/80 120 Pr',ary System:

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Water Volume (ft 33) 12,710 11,347 11,671 13 40 78 Steam Volume (ft ) 720 720  ?

222,739 222,739

+ Mass of U0 2 in Core (1b) 216,600 216,600 Mass of Steel in Core (1b) 21,000 ' 20,407 7 19,000 Mass of Zr in Core (1b) 44,500 44,600 45,296 45,234-Mass of Bottom Head (lb) 87,000 78,130 87,000 87,000 Bottom Heid Diameter (ft) 14.4 14.7 14.4 14.4 Bottom Head Thickness (ft) 0.45 0.44 0.45 0.45 Containment Building Coolers:

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

~ Total Mass of Water (1b) - -200,000 173,000 348,000 213,000 Initial Pressure (psia) 665 665 600 615 Temperature _ (*F), ,

150 150 80 $ /c'2 -/ O Refuelino Water Storace Tank:

Total Mass of Water (Ib). 2.89 x 10 8 2.89 x 10' 10 7 2.89'x 10s.

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

Configuration Wet Wet Dry

• Dry / Wet Concrete Material Limestone Basaltic Basaltic Basaltic

• Minimum (Maximum Capacity = 3.9 x 10' lb) l .

3 I

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a i

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

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3. - ASSESSMENT OF CONTAINMENT PERFORMANCE in this chapter, the review of contairnent responce to severe accidents is described. Analytical techniques used to analyze core meltdown phenomena 4 and containment response are reviewed, containment failure model is assessed and plant damage states and containment failure modes are evaluated.

Parallels between this study and other PRAs are cet forth. Finally, 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 meltdown and containment response analyses is provided in this section.

The MARCH 8 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 .

pressure until the vessel failure (" time-phased dispersal"), the . H00 MESH 5 r

coriputer code is used. This code calculates the steam and hydrogen blowdown

! from the reactor vessel using an isothemal ideal gas model. The water level boil-off from. the reactor cavity 'loor is modeled using a saturated critical heat flux correlation.. AdditionCly, the accumulator discharge following de-pressurization caused by the vessel failure is also considered.

A modified version of the CORCON' code is used to replace the INTERS sub-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 i proper sequencing and integration into- the overall mass and energy input to COC0 CLASS 9s code.

C0C0 CLASS 9, a modified version of the Westinghouse COCO computer code utilizes the mass and energy inputs to the contaiment as computed by MARCH to 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-ment heat removal through contalment sprays and sump recirculation.

Fission product transport and consequence calculations are perfomed using the CORRAL-Il and the PLG proprietary CRACIT 5 computer codes, respec-l tively.

f The analytical methods used. to carry out the core and contaiment themal l hydraulics, and fission product transport calculations are identical to ttiose l used for MPSS-3.7 ,

l l

i .

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 couoted with the status of reactor cavity and the spray system, determine the source terns. This section deals with the mode and time of containment fail-ure.

- 3.2.2 Design Description The primary contaiment of the Seabrook plant'is a seismic Category I re-inforccd concrete dry structure. It consists of an upright cylinder topped with a he aispherical dome. The inside diameter of the cylinder is 140 feet

and the'inside 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 106 ft 3. ya The[inside of the containment 1sjwelded yCr.hM' steel liner. 'The liner plate it the cylinder is 3/8" thick in all areas except penetration and the b tion of the basemat and cylinder where it is 3/4" thick. This liner i 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 pemits leak

testing of these welds throughout the life of tha 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.

1 The containment building is surrounded by an enclosure. The contair. ment 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.

l 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 contalment enclosure is designed to control any leakage from the containment structure. To accomplish this, the space between the containment and the enclosure building (approximately 4'6" wide) is maintained at a slight i

negative pressure (-0.25" water gauge) during accident conditions by fans which take suction from the contalment enclosure and exhaust to atmosphere through charcoal filters. '

There are a number of containment penetrations which are steel components that resist pressure. These penetrations are not backed by structural con .

crete and include.the following:

P CONTAINMENT TIME OF FAILURE FAILURE MODE WET OR DRY RELEASE SPRAY REACTOR CAVITY CATEGORY SYSTEM SOURCE TERM Figure 3.1 A schematic representation of source term calculation.

O e

Table 3.1 Containment Operating and Design Parameters Parameter Value Normal ~0peration .

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 P'ressu.e , inches w.g. -0.25

- Design Conditions Pressure , psig 5'2.0 Temperature , F 296 Free Volume , ft3 2.7x106 Leak Rate , % mass / day 0.[

Containment Enclosure Pressure , psig -3.5 j) e..e UA p) 'l '"%

l '

0 2

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 containment enclosure shells to provide access, ar.chor piping, or furnish some other opera.ional requirement.

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

Equipment Hatch The equipment . hatch (Fig. 3.2) consists of the barrel, the spherical dished cover plate with flange, and the air lock mounting sleeve. The center-line of the hatch is located at elevation 37'1/2" and an azimuth of 150*. The 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 (inchesh Barrel 3 1/2 Spherical 1 3/8 Flange 5 3/8 Air lock mounting 'l 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-tached to the hatch barrel with 32 swing bolts,1-3/8 inch in diameter. The barrel, which is also the sleeve for the equipment hatch, is embedded in the

. shell of the concrete containment. The equipment hatch 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: *

' 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 8

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9

13-are located such that the area between doors can be pressurized to 52.0 psig.

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- .nel Ai r Lock Th'e 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 lock, is imbedded in the shell of the concrete contairnent. The centerline of the carrel 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 F1ange on Door 7' 9 1/8" Barrel 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 contairnent. 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 locking device for the doors is a rotating, third ring, breach-type mechanism. These doors are also mecnanically interlocked 50 that only one door can be opened at a time. The capability exists for bypassing this interlock and relieving the internal pressure by use of special tools.

The doors may be operated mechanically. .

Piping Penetrations There are two types of piping penetrations: moderate energy and high e ne rgy. Moderate energy piping penetrations are used 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.

l

, 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 support inside the containment. The sliding support provides shear restraint l

while pemitting relative motion between the pipe and the support. The annu-lar space between the process pipe and the sleeve is completely filled with fiberglass themal 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 l concrete contaiment, ASME III (CC). The sliding support inside the contain-ment is classified as an ASME Safety Class 2 component support (NF). '

Moderate energy piping penetrations (Fig. 3.5) consist of one or more process pipes, the contaiment penetration sleeve, and a flat circular end .

pl ate. 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.

Electrical and Instrumentation Penetrations Electrical pennrations (Fig. 3.6) consist of a stainless steel header '

plate with an attach'ed terminal 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, stainless steel header plate and carbon steel weld ring were designed as ASME III Safety Cla'ss 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.

-Instrumentation penetrations are of two types -- electrical and fl uid.

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 assembly consists of the fuel transfer tube, the penetration sleeve, the fixed sat.*e on the reactor side, and the sliding sad-die in the fuel storage building. The fuel transfer tube centerline is at elevation (-)9'4-1/4" and it has approximately 20" inner diameter. The 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 contalment air purge penetrations (HVAC-1 and HVAC-2) and the containment on-line penetra-tions (X-16 and X-18h The contaiment air purge penetrations (Fig. 3.7) each consist of a pipe sleeve (a rolled and welded pipe section, 36" outer diameter by 1/2" wall thickness)' which is flanged at each and with 36" weld neck flanges and, attached to these flanges, the inner and outer isolation valves.

Together with the pipe, these valves fom a part of the containment pressure bounda ry. The valves are 36" diameter butterfly valves with fail-safe pneu-matic ' operators. The. weld between the pipe and

• the contairstent liner is equipped with a leak chase for pressure testing. j The containment on-line purge penetrations each consist of a pipe sleeve (a rolled and welded pipe section, 8" o.d. by 1/2" wall thickness). A short section of pipe with a nipple is welded to the sleeve on the outside of the-contaiment, and a 3/4" valve and test connection is attached to it. The

. 18-Table 3.2 Containment Liner Penetrations a

Penetration .

Penetration Numbers Service Size X-1 to X-4 Main steam line 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 inj ction 8" X-14 to X-15 Containment building spray 8" X-16, X-18 Containment on-l'r.e 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 injection 2" X-32, X-34 Orain line 3",2" '

X-33, X-37 CVCS 3" X-35, X-36, X-40 RCS test / sample control X-52, X-71, X-72 1" or smaller X-38 Combustible gas control 10" X-39 Spent fuel pool cooling 2" t X-43, X-47, X-50 Instrumentation lines

X-57 d)' <f (

I X-60, X-61 From containment rectreulation sump 16" l X-62 Fuel transfer tube

• 20" X-63 to X-66 Steam generator blowdown 3" X-67 ' Service air '

2" j HVAC-1,2 Containment purge supply / exhaust lines 36" X-19, X-41, X-42 X-44 to X-46, X-48 ',-

Spare ,

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i X-49, X-51, X-58 X-59, X-68 to X-70 i .

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Figure 3.6 Tyical electrical penetration.

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ISOLATION VALVES ENCLOSURE BUILDING CONTAINMENT WALL

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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.eakaoe Rate Calculation Under severe accident conditions the pressure inside the containment quickly builds up in the range of 75 to 200 psi. At these pressures, any leakage through the contaiment holes will essentially be choked. The .] I leakage under choked flow condition is given as (Ref.10):

k+1 W= k(k 1) A III '

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 p=g (2) where R = gas constant, and T = the absoluta temperature, Then Eq.(1) becomes k+1 k-1 -

W= k(k+1)

A (3)

The mass of mixture can be written as M = Yp or, H=h , (4) where V is the free mixture volume in the contaiment. Equations (3) an (4) can be combined to get the leakage rate, in terms of mass fraction, as e

o k+1 f=%k(g2)M vTT A ,

(5)

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

i For Seabroot-1, using V = 2.704x106 ft 3 and T = 296 F, Eq.(5) gives Leakage Rate =-0.721 Ai n 'W/o Per hour (6) where Ai n is the leakage area in in2 . Al te rnately, 1 Leakage Rate = 17.3 Ai n W/o Per day. (7)

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

about 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to leak the entire content to the enviroment through a 10-in 2 hole.

)

3.2.4 Containment Failure Model -

3.2.4.1 Leak-Before-Failure .

l During accident sequences involving core damage, the containment struc-

! ture will be exposed to pressures and temperatures beyond those used in the I

design basis accident (DBA). Response of the contairunent building to these

, severe conditions is evaluated in SSPSA by employing, for the first time, a

! leak-before-failure model. In this model allowance is made for continuous l 1eakage from the containment to the surroundings. This mode of contaiment

failure is tenned 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 i

ar.d 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-fall'ure model, the release of activity, which is considerably small compared with that for the gross fall,ure 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.

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

3.2.4.2 Class 4fication of Failure The SSPSA report has clas,1fied containment failures in three categories:

-. Conta'nment Failure Category A. Includes containment failures that develop a small leak that is substantially larger than the leak ac-

ceptable from an intact containment, but not large enough to arrest ,

lthe pressure rise in the contairrnent. Category A failures thus cause an early increase in the rate of leakage of radionuclides over the de-sign basis leak rate but pressurization of the contaiment continues '

until either a category B or C contalment failure occurs.

i The ited by intact t5e contairunent Technical Specification is definedvalue. as theFor one in which leakage Seabrook-1, this value is lim- gC 6

is 0.2 w/o per day at the calculated peak accident pressure of approx /

imately 47 psig. Note that the SSP has used 0.1 volume perf- p v

cent per day for this leakage, al though/ prior to Ine most recent i amendment dated August 1984, the 4; L.2 cited 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 j 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 containment no longer increases. The time during which a substantial fraction of the radionuclide source tenn is released is longer than approximataly 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Category B failures include self-regulating failure modes where the leak area is initially small but increases with pressure so 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 contalment failure i

modes that develop a large leak area. A large fraction of the total radionuclide so~urce term is released over a period of . lese-then 11. _ /,

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

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

, / ADBA MAA ANP Type A Ag ' < Ap Type B (8) g.fme AC Type C ,

where. .

l' f

gi /g-ADBA = leakage area corresponding to th,e technical specification limit for contaiment leakage, ,

! *Tnere appears to be substantial update /chan9es 'in theAmyir.eered Safety -

j Features flow.Alegram, including,arr3ng ments of moto(opera ed valves and bypass lines', which may substantially ch nge the, frequency o events. BNL,.

[

w howeverjis not reviewijg-this part of SS SAf.

e j

i

- - - - - - .n,n._ - , , - - - - - - - - - , - --- _._..,,,,-- ,--- _ -- .---,-- - - - - , - - . - - - - - - - , - , - - -

• f 3, L (. Anp = leakage area not large enough to arrest 3

j

, j .\ g pressurizat, ion, and

\ Ap = leakage area sufficient to release 100 w/o

, f

. t in one hour. l v .1 The leakage area required to release a subst al fraction of the radio- '

j W inuclide source tem in approximately an hour @> can be computed using Eq.

1

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

', p .A 1 ft . Therefore, any containment leak area in excess of 1 ft 2 will be -

Q k .hfined as a gross _.contaiment failure-(Catecorv.C). This estimate of the ea '

^ area is factor two too high from the value stated in ss ~^ #

W j The leakage area required to arrest containment pressurization is in the I range of 4 to 10 square inches, the lower value being more representative of-wet sequences and the upper value is representative of dry sequences. A leak area of about 6 square inches will result in the release of about 100 w/o of ac-ivity in a day (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

in about 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The Category 8 leak area is, thus, in the j (100%ofturnover}

range 4 in to 21 ftFigure

. 3.8 is a pictorial representation of these ,

leakage categcries.

3.2.5 Containment Pressure Capacity ,

3.2.5.1 Concrete Containment T1e Seabrook PSA has examined failure modes for the containment structure itself, the steel liner, all penetrations, equipment and personnel lock hatch-es, and the secondary antainment. The containment structure includes the cylindrical wall, the hemispherical dome, the base slab and the base slab and i containment wall junction. The most critical membrane tension was found to l

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-

imately 157 psi, with a coefficient of variation of 0.084 The ultimate hoop l

load in cylinder is 216 psig. The contaiment wall is, tnus, tssumed to fail at this pressure. At pressures beycad this, very large irreversible defoma-i tions occur which will cause cracks in the reinforced concrete but the loss of integrity of the pressure boundary may not occur until the liner tears. The

' compiled radial defomations of the containment wall are shown in Figure 3.9.

Note that the radial strain at the expected failure pressure of 216 psi is l 4.77. (ar/r).

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

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

failure is more restrictive. For this mode, the median failure pressure is i

estimated in SSPSA as 323 psig, with a logarithmic standard deviatio'n of 0.23. Although the uncertainty for failure of the base slab is large, the probability of failure is small because the median capacities are high. Thus, failure of the base slab is not considered to be a critical failure mode and

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......e.e.e -. e-e-e-..e.

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e i Figure 3.8 A pictorial representation df leakage categories. ' ' "

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

failure.

~

Secondary stresses in the cylindrical portion of the contaiment occur at discontinuity such as at the base slab contai nment 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 contalment wall junc-tion is not limiting.

In sumary, 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) failure occurs.

3.2.5.2 Linar 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 untaxial 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 concentrations. None of the meridional or hoop reinforcing bars are teminated at penetrations. Instead, I

they are continued around the penetrations,'thus ensuring that excess hoop and j meridional capacity is available. Table 3.2 lists all piping penetrations, j 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 contaiment wall and the pipe whip re-i straints. Therefore, the most critical penetrations are the areas where the J

pipe is supported close to the penetration. Also, stronger and stiffer pipes 1

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

I 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) ~

l Penetration X-26 4" schedule 160, stainless steel 3 (also X-24, X-25, X-27) 1 3

l

)

3

_______h., _ . _ _ _ , . ____ _..._ ___

Penetration X-71 1" - multiple pipe penetration i (also X-72 and possibly*

others)  ;

> Penetration X-8 18" ~ main feedwater schedule 100, (also X-5 tc X-7) carbon steel l

~

j l'uel Transfer Tube Convoluted Bellows The probability of failure at these penetrations was computed by (a) establishing a pressure-displacement . relation, (b) estimating ther failure probability as a function cf radial displacement and then (c) combining' the 4

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 defomation of l contaiment would also be small. -

1' .

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 use 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.

The dian f ailure ressure for X-23 penetration, t k es _ 2

! n n s e is higher than the hoop failure ressu e (216 psig) of a r

e contaiment wall. These leak areas, therefore, are not expected to devel-l op.

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

The combined leak area for all safety injection penetrations is obtained by

independently adding individual median leak area of 0.5 inz, i

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

The main fecdoter 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

! he d due to axial loads in the pipe at the penetration. At a median pressure o 180 psig,2each one of these penetrations is likely to result in a leak l

area of 50 in each. Since all four of these can fail independent of each other, the total leak area is 200 in . 2 Although the failure of a single such penetration can be considered as Type B failure,11; all four main feedwater l

penetrations were to fail simultaneously the resulting leakage will be of Type f

! C. ~

l The fuel transfer tube is fixed to an elevated floor inside the contain-As the pressure in the containment increases, the containment wall.

ment.

moves outwards and thereby exerts pressure on the bellows. The most pertinent i

\

l l

l

Table 3.3 Leak Area Estimates for Mechanical Penetrations Median Median Line Penetration Leak Area Failure Pressure Size in2 p3gg X .20 to X-23 6.0 >216 12" CCW Supply and Return X-24 to X-27 2.0 166 4" Safety Injection X-71 and X-72 Negligible < 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 O

1 e'

I 1

i .

O 1

4 l

0 o

. . - i bellows from the viewpoint of containment leakage is the one inside the con-tainment (EP-2). Three potential f ailure 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) f ailure due to meridional bending strains. The SSPSA has estimated median leak area of about 3 in2 at a pressure of about 172 psig. This is a Type A failure., ,

There are two sets of containment penetrations which are cpen to the contal ment atmcsphere on the inside. The on-line penetrations (X-16 and X-18) are the 8-inch purge suction and discharge lines and containment 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 contaiment. All eight valves are pneumatically operated butter-fly valies. At elevated temperatures, the seal material (usually ethylene propylene) on thes valves may deteriorate and lose ' its sealing function.

Any deposition of radioactive aerosols could further deteriorate the sealant material. Consideri .g sealant degradation due to temperature alone, ethylene propylene seal life (Ref.10) is 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, 40 mts, or 20 mts if exposed to 400, 500 or 600 F, respectively.

In the event of the failure of the sealant m ei ,an crack leak.

path may develop :nd containment atmosphere may en to leak into the space between the two isolation valves. Since the isola on valves are closed from the containment isolation signal system, the leakage of containment atmosphere to the enviroment 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 (P.ef.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 int as low value and 34 in 2 as high value. As noted ear 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 containment failure (Category C).

Electrical penetrations can fail primarily due to overheating of the pot-ti ng compound. Tte SSPSA study has concluded that the failure of electrical pe etrations is-not expected to make a significant contribution to containment failure for any accident sequence. This conclusion, appears justified for the  ;

wet case, but, for the dry case, it is based on their estimate of slow over-  ;

heating of the potting compound. A careful themal' conduction calcula~ tion should be made to check this assessment. Such a calculation, similar to the problem of vent / purge line butterfly isolation valve failure, is beyonde 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 sill-co ne) . The structural f ailure, prior to containment f ailure, appears unlike-ly. The sealant material can degrade at high temperatures typical of a

! . . . j severe accident. According to the 0-Ring Handbook (see Ref.10), silicone 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-ment hatch and personnel lock penetrations do not contribute significantly to Type B 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 since anticipated leak area is 50 in .

2 It appears to us that when one such penetration fails, the remain-ing three will also fail 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 prc3sure substantially prior to the failure of the three remaining penetrations. Assuming that all four main feedwater lines fall at 180 psig, an equivalent leak area of 200 in 2 will result. This fail-ure, therefore, should be classified as Type C. The impact of this change on the - containment failure probability numbers will be to reduce the rate for .

Type B with a corresponding increase in Type C. The total failure rate is not i likely to change. Estimated containmant failure fractions are compared with

! the SSPSA results in Fig. 3.10.

3.2.5.5 Containment Enclosure The containment enclosure building is designed to withstand 3.5 psipres-sure difference between the enclosure and the environment. During nomal operation, the internal pressure is about -0.25 inches of water gauge. The SSPSA study has calculated its pressure capacity to range from more than l

1 psid to 10 psid. In view of relatively strong primary containment, the role of the secondary containment is important primarily for Type 8 failures of the i primary contaiment. In the event of Type C failure, the secondary enclosure building might not play any significant role as far as the source term calcu-lation is concerned.

l 3.3 Definition of Plant Damage States and Contaiment Response 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 of all the individual scenarios in that category.

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

• ~

1. Initiating Events

" A" - Large Loss of Coolant Accident "S" -

Small loss of Coolant Accident *

"T" -

Transient e

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e. e e.. .

O

~

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

}

' ( ,

"F W" - Emergency Feedwater

3. Availability of Containment Systems "C" - Long-Term Contaiment Spray Cooling

! "4" - Long-Term Speay Recirculation, No Cooling

! "I" - Isolation Failure er Bypass

> 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.

1 These damage states are categcrized in a matrix of eight physical conditions in the containment (numerals (1) to (8)) and six combinations of containment safetyf 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 i sequences involving steam generator tube ruptures. Figure 3.11 indicates that 4 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 contaiment class is the sum of the frequencies of the plant .

states included therein.

Annual plant state frequencies calculated by the applicants for both in-ternal and external events were reviewed by the Lawrence Livemore National Laboratory' and were found acceptable. Table 3.6 presents, the calculated i contaiment class frequency estimates for internal events, fires, floods and j truck crashes; moderate and severe seismic events.

In order to comprehensively assess the risk from seismic events, it is necessary to make separate consequence calculations for those accidents which j are initiated by earthquakes severe enough to impair evacuation. For this

• purpose, the seismic frequency estimates are divided into two categories in 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 events'with acceleration greater than 0.50g are judged to impair evac-l l

i untion, and must be treated separately in the consequence analysis. Ml i These contaiment response classes (or ant damage states) are the i starting point for the containment event bree nalysis and they define the i link or interfaces with the plant analysis. g l 3.4 Contalment Event Tree and Accident Phenomenology 1

An important step towards the development of the contaiment matrix in-l volves the quantification of branch point probabilities in the containment i event tree. These probabilities depend heavily on the analyses of degraded-and core melt phenomenology and the contalment building response described in

, Appendix H of the SSPSA.s 1

i I .

f

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Plant State Represents 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 AEC1 4E/8E TEC1 IF V 2F/6F AEI 3F/7F SEI 4F/EF SLI .

Figure 3.15' Definitions of the plant darnage states used in SPSS.

Table 3.4 Frequencies of Occurrence of the Plant Dama9e States Frequency Fruquency Plant Damage (eventsper Plant Damage (evente per State reactor year) State reactor year) 9 10 .3.03(-7) 6A 3.41(-7)

IF 1.89(-6) 6C 3.57(-10) 1FA 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 notaticn is indicated in abbreviated form; i.e., 3.03(-7) = 3.03 x 10-7 -

- 0 e

l

=

e

36 Table 3.5 Containment Response Class Cefinitions Class Plant State Represents 1 10 AE 2- 2A/6A, 4A/8A AEC, TEC, SEC 3 2C/6C, 4C/8C AE4, TE4, SE4 ,

4 30/70 SE, TE, TEFW 5 20/60, 40/80 AL, SL, TL 6 -

IF, 2F, 3F, 4F, 6F, y 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)

I

+

5 l

• h e

. . . . . . - - . . _ , . . . _ - . . , , - _, _.,-.._,-----.__n,-, .

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

, Containment Internal, Fires, Internal

! Response Class Floods and Truck Seismic <0.5g Seismic >0.59 Total Seismic and l

Crashes External I

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 2

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 2.4 E-6 -

7 . . . . .

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 * - - - *

• 11 * - - - * ~

12 - - -

• 13 * - - -

• l tReference [5] Tables 5.1-3 and 9.2-9.

I

• Indicates frequencies less than.10-8 yr 1 j .

l og #

f 2 ,

- . s

-- . - . .-- - . - - .- . .. - ~ . . - . - . - - . . - - . - -

! a38-i .

The SSPSA contalment event tree uses the twelve top events identified in j Table 3.7 as major- phenomenojogical phases which could occur with respect to  !

! the formation and lo,ca,t core debris. These processes are grouped into i four phases folloyt g of,ident' (c initiation (1) phenomena occurring while the core is still n (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-taiment floor; and (4) the phenomena involving long-tem coolin con-j tainment and/or basemat penetration.

) 3.5 Containment Matrix (C-Matrix)

n. -

Tne twelve top event.s in the Seabrook contalment event tree are summar-i ized in Table 3.7. A negative response at any of the five nodes (4. 8,10, 1

.'11, i taiment andbuilcing

12) in the by acontaiment event variety of failure modes. tree results in the failure of the con- [l 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 contsiment event tree thus links the plant damage states to a range of possible containment failure modes via the various paths through the tree. For a given tree, each path l ends in a coditional probability (Cp) of' occurrence, and these cps should sum

to unity. The quantification of an event tree is the process by which all the ,

j paths are combined to give the conditional probabilities of the various

! release categories. In SSPSA, fourteen release categories are used for the i ,

quantification as . summarized in Table 3.8. Note that two of these release P categories (namely, SS and 53 ) correspond to intact / isolated contaiment.

j' Fission product release for this category would, therefore, be via nomal

leakage paths in the containment (ard enclosure) building, which can be if= ~3 ,

j farent-depending on availability of the enclosure building v'entilatio fil- -

ation system.

Table 3.9 sets forth'a s!mplified containment matrix (C-matrix) for the 3

Seabrook plant using the contaiment response class definitions discussed in 1 ,

Section 3.3, and the release category definitions given in Table 3.8. In

• l~

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.

I

~

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

} risk studies for other 1 ,3," , pressurized water reactors with large dry 1 . containments is used to guide this assessment.

The mode and timing of contalment failure cannot be calculated with a j great degree of accuracy. Judgements must be made about the nature of the '

i -

dominant phenomena and about the magnitude of several important parameters.

[ . .. Furthemore, the codes and methods used for these calculations are approximate i* and do not model all of the detailed phenomena. Fortunately, risk measured in i personal exposure is not sensitive to minor variations in failure mode /and.

^

timing. It is important, however, tc prcperly cha acterize the major attri-butes of failure mechanisms; (1) whether the failure is early or late, (2)

I whether it is by overpressurization, bypass, or basemat melt-through and, (3)*

{

whether or not radionuclide removal systems are effective.

i

.w-- a - - . . n. -e--n..n__-,,-_+,,,,.,,-a-,,n, n____,,m,wn,ma--nn,_ _,n, , _,-.p.3-,,-,,,,.,-,

Table 3.7 Accident Phase and Top Events for the Seab, rook Containment Event Tree Accident Phase Top Event Initiator 1 Plant State D'ebe . in Vessel 2 Debris Cooled in Place 3 No H2 Burn ,

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) -

S O

I e

i

,-._-..m,._. - __.

O e

-40 Table 3.8 Release Categories Employed in the Seabrook Station Risk Model Release Category Release

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

Intact / Isolated .

SS Same as S5 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.

S2V Same as U, but with an additional vaporiza-tion coepenent of the source term.

53 Late overpressurization failure of the con- -

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

Failure -

U Same as S3, but with containment building sprays not working.

S3V Same as U, but with an additional vaporiza-tion component of the source term; S4 Basemat penetration failure, sprays operating S4V Containment basemat penetration failure with containment building sprays not working and additional vaporization component of the source term.

56 Containment bypass or isolation f ailure with containment building sprays working.

S6V Same as 56, but with containment building sprays not working and an additional vapori-Early zation component of the source term.

Containment-Failure / Bypass S1 Early containment ~ failure due to steam expl'-

o sion or hydrogen burn,with containment building sprays working. ,

U Same as S1, but with containment building

~

sorays not working.

• 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.

e

41 Table 3.9 Sieplified Centainment Matrix for Seabrook Release Category Class 51 52 53 55 56 57 TI S2V S3V S4v 56v S5V-d 1 .

0.60 0.40 2 0.01 0.99 3 1.0 4

0.89 0.11 5 1.0 6

1.0 7 1.0 8 1.0 9 1.0 t

l 10 1.0

11 1.0

• 12 -

1. 0 13 1.0 O

e O

M O

e

--- __,y- _ , - _ . . _ _ . ,

=. _- .. . . _ . .

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

Early Failure (51, H) which result fra a steam explosion or an early hy-crogen burn is oelieved t unlikely. Altnough explosions in the reactor vessel lower' plenum are probable, the resulting mechanical energy would .

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

The conditional probability for an early contalment failure due to ex-ternal e nts (i.e., aircraft crashes) is assigned 1 in the SSPSA as shown in

! Table . This simply indicates that an aircraft crash into the contaiment is assumed to fail the containment structure with certainty.

"$3d J. T (

Early containment failure could also conceivably result from direct heat-trig due to a rapid dispersal of the core debris throughout containment in the fom of aerosols. The dispersal could only be caused by the high primary sys- -

tem pressures- that may exist at vessel failure for a number of transient se-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 could rapidly pres-surize containment by direct heat exchange and concomitant chemical reac-tions. Scoping calculations perfonned by the Containment Loads Working Group (CLWG) showed that a very severe challenge to the contalment integrity could result provided 25 percent of the core mass were converted to aerosols.12 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. -

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

For the reasons outlined above as well as the high containment failure pressure for Seabrook, it is concluded that early overpressure failure has a very low likelihood.

h Early Containment Leakage (S2, 37, TflT) without gross failure of containment building is expected to occur 3r nonisolated steam generator tube rupture

, , event with containment sprays available. (S2), for large break LOCA sequences with RWST i njection in the absence of sprays ('51) , and for dry cav.ity sequences with a vaporization release (WiI).

There seems to be a basic inconsistency in assigning plant damage staftes to this failure mode as defined in the C-matrix. Specifically, large break ~

G LOCA sequences with RWST injection in the absence of containment sprays are expected to lead to an D failure mode with 100% probability (see U below); .

+

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

while they are also assigned to 37 with 100'. probability. This can be correct only if the initiator and the sequences are indeed different, but at this time we cannot resolve the inconsistency.

Similarly, the significance of contaiment functions on steam generator g tube rupture sequences is not at all obvious.

1. ate Overpressuriza*. ion Failure (53, 57, ITI) can occur due to steam produc- -

tun in a wet cavity or no:.concensable gas production es a result of core-con-crete interaction for a dry cavity situation. For sequences in which early and intermediate failure is not expected to occur, and for which contaiment sprays are inoperabic, failure is expected to be a certainty.

The conditional probability for a late overpressurization failure with a vaporization release (dry cavity) is shown to be 0.60. This results from the relative competition between the late overpressure failure and the basemat penetration (3W) for accident sequences without the containment sprays.

The failure time for the late overpressurization failure mode is much longer than previously calculated for other large dry contaiment.I s3e" This is as a result of the very high failure pressure for the Seabrook con-tai me nt. As a consequence of this high contalment 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.

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 contaiment 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, 3TV) can only result in the absence of con-tainment heat 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 conservatively assigned.

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

l Ine raolological releases are thus limited to the design basis leakage with essentially negligible of f-site consequences. ,

Containment Isolation Failure (56, 5TV) is represented by an 8-inch diameter

pu rge line. Ine accident sequences where the containment is either enot , -

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

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

and, at failure, the temperature can exceed 700*F.

e

. . . j l

• isolated or bypassed (Event V) are assigned a conditional probability 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 contain-me nt . Furthermore, the concurrent assumed loss of RHR and coolant make-up

• capabil-ity leads to severe core damage. In the SPSS, three possible inter-facing systems LOCA sequences have been found and discussed. These are

1. Disc rupture of the check valve in the cold-leg injectior$ itnes of the RHR. .

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

3. Disc rupture of the motor-operated valve equipped with a steam mount-ed limit switch and " disc f ailing open while. indicated closed" in the other motor-opercted valve in the normal 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 containment event tree and thus the C-matrix are in general ag reement with the other previously NRC reviewed studies 1 .3.4 for PWRs with large dry contaiments. Furthermore, the claimed unusually-high strength of the Seabrook contairnent reduces the impact of sensitivity ' caused by uncer-tainties in the severe accident progression. However, should the claimed stre gth of the contalment 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 Cateoory Frecuencies Based on the cantaiment class frequencies in Table 3.6 and the contain-ment failure matrix of Table 3.9, the ' release frequencies were computed and i are summarized in Table 3.10. g,/f

, tf t

Table 3.10 indicates that on1 light f the release categories dominate the total release frequency.

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

$3V dominate the sou c tem frequency. j

\

l j .

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 53 7.50E-7 3.45E-8 2.69E-7 1.05E-6 S5 5.64E-5 1.52E-6 1.23E-6 5.92E-5 37 -

• 1.12E-7 2.40E-7 3.52E-7 37 5.50E-5 1.10E-6 1.76E-6 5.79E-5 S2v 5.29E-6 1.25E-5 1.78E-5 STI 7.66E-5 1.65E-6 2.14E-6 8.04E-5 54v 9.50E-6 2.04E-7 3.27E-7 1.0E-5 T57 ,

1.80E-6 1.66E-7 3.93E-7 2.36E-6 I

e o

e I

i

Table 3.11 Contribution of Containment Response Classes to the Total Core Melt Frequency Internal, Fires, . Internal Containment Floods and Truck and Class Crashes Seismic <0.5g Seismic >0.59 Total Seismic External 1 - - - -

<0.01 2 0.25 <0.01 <0.01 0.01 0.26 i

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 i  ; .

J * * * *

• i 7

! 8

• 0.025 0.055 0.08 0.08 9-13 * * * *

• W*

e l

l

Table.3.12 Release Category Frequency as a Fraction of Core Mcit Frequency Release Internal, Fires, Internal Category Floods and Truck .

and Crashes Seismic <0.59 Seismic >0.59 Total Seismic External

. 53 <0.01 i

<0.01 <0.01 <0.01 ' <0.01 SS .

,

• 0.25 <

<0.01 <0.01 0.01 O.26 S2 <0.01 <0.01 <0.01 <0.01 lei 0.24 <0.01 <0.01 0.01 0.25 ,

S2V 0.03 0.05 0.08 0.08 l[UI 0.33 0.01 0.01 0.02 0.35 S4V O.04 <0.01 <0.01 <0.01 0.04 f:

56V 0.01 <0.01 <0.01 <0.01 0.01 I

a 1

4. ACCIDENT SOURCE TERMS In this chapter the approach utilized in the SSPSA to detemine tihe fraction of fission products originally in the core which can leak to the out-side environment will be outlined. The fission product source to the environ-ment as calculated by this approach for each release category will also be discussed. ,

4.1 Assesscent of Severe Accident Source Terms As in the Reactor Safety udy (RSS)u the CORRAL-II code was used in the SSPSA for detemining fission pro?uct leakage to the enviroment. This code k

takes input from the thermal-hydraulic analysis carried out- for the contain-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 time dependence of these phases is determined by the timing of core heatup, primary system f ailure, and core / concrete interaction. The methods used in '

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 g roup ,

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

3) An uncertainty analysis was carried out in which it was attempted to account for shortcomings in the RSS methods.

In general, the net result of the SSPSA analysis was to reduce the fractional release of particulate fission products. This will be discussed in more de-tail later. -In all, fourteen releases were detemined ranging from contain-ment bypass sequence to the no-fall sequence as shown in Table 3.8.

These release categories were evaluated ' by considering the containment failure mode, the' availability of the spray system, and whether or not the cavity was wet or dry. Table 4.1 shows the point-estimate' releases as deter-mined by the methods outlined above. Containment. failure mode S1 corresponds to a gross failure of the contalment, resulting from a steam explosion, early pressure spikes, or early hydrogen burns. Failure mode 52 represents a loss of contaiment function early in the accident sequence. This loss of function takes the fom of an increase in the leak rate to 40% pcr day where it stays until the containment fails due to overpressurization. Failure mode 53 repre-sents a late overpressurization failure of the containment driven by decay heat or late hydrogen burns. Failure mode S4 represents a basemat mel t-through, 55 represents no containment failure and the leak rate is limited to the contaiment design basis leak rate. Finally, failure mode 56 represents i sequences where the containment is failed or bypassed as part of the initi'at-ing event.

The second parameter considered in defining the source term is the avail-ability of sprays. This is determined by the plant damage states. Those'

- - ~ , -

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e w ar e w w e .= .=

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• w ==OON m c"* * #*M en g .* N "O e* O ==*=*ar OO r#

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o e

9

release categories with operating spray systems are designated $1 to S6, while those with spray systems,not operating are designated Tf to T6.

The third and final parameter considered in differentiating between source tems 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 "/" in their designation as shown in Table 3.8.

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

Release categories 37 and W have late overpressurization failure 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 wet sequences and 187 psia for dry sequences. At this pressure a gross failure is expected result-i ng in a puff release of approximately 0.5 hr release duration. From Table 4.1 it is seen that the ST and 53V sequences fail at 27.2 hrs and 91.5 hrs, respectively. These failure times are several hours later than was cal-culated for Indian Point, Zion, and Millstone-3. The primary reason for the later failure in this t.ase is due to the superior strength of the containment st ructu re. Table 4.2 compares the T3, SW release parameters with similar 1

parameters for the o-her three reactors mentioned - above. Note that a fair comparison should set (0!+1) equal to (Cs-Rb), since iodine was treated as Csi. 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 approximately an order of magnitude. This difference is due to the latter failure 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 53.* 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 radionuclide transport mechanisms inside the contain-ment due to changes in the containment themal hydraulic conditions.

Release category 3R is associated with early containment failure in which the containment 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 multi-puff ' release. The first puff corresponds to the releg up to the time when vaporization Starts i (melt +ga2). The secon puss ncludes the period of vaporization release ,and I the third puff is equi to an overpressurization failure at the time of catastrophic containmer; failure. In this model -the duration of the melt  !

(

Table 4.2 Late Overpressurization Failure Comparison Millstone-3 7 Zion / Indian!" Indian 3 Seabrook5 Point Study Point IT 3TT H-7 TMLB' 2RW Xe 9.0(-1) 1.0 9 (-1) 9.6(-1) 1.0 10+I ~ 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)

Te-Sb 2.2(-2) 3.0(-2) 3.0(-1) 3.8(-1) 4.4(-1)

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)

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)

' Energy 300E7 300E7 540E6 150E6 (Btu /hr) 9 f

.G release is s n to be 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. vaporization release 7.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the re-mai ni ng ase 78.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. It is not clear that the melt release in this case is 5.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, however, it does not seem to be unreasonable. A 7.2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> duration or the vaporization release is not consistent with the RSS,13 which  ;

only allows 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for this phase. Finally, it is not clear how the 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> l for the last phase was deten:,ined. The release duration for a single puff, l which is ,he sum of the above three phases leads to a release time of 88.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> whic" 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 Millstone-3 study. This comparison is made in Table 4.3. It is seen that, once adjustments are made for the dif-ferent ways in which iodine is treated, the 5F release.is approximately half the M-4 rel 2ase. Without .the benefit of a calculation, it is difficult to judge whether the differences are reasonable. However, a possible reason for this reduction is the credit taken for the enclosure building surrounding the /C actual containment building. This feature is unique to the Seabrook contain-ment structure. .

Release category T67 has binned into it an isolation failure corresp;nd-ing to an 8" diameter breach in containment and the interfacing LOCA (Y-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 RSS 13 consequenca mo al. Our recommendation would be to reduce these times to more reasonable values (see Table 4.7). - /f The release fraction can be compared (Table 4.3) to the M-4 release from Millstone-3, PWR-2 for the RSS and the V-sequence from the RSSMAP study for So rry.15 Except 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 the value for S6V would be the lowest release fraction.

4.2 Source Term U.; certainty 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 -

l

2) Uncertairties in timing of key events.

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Table 4.3 Comparison of Releases for Failure to isolate Containment and the By-Pasi Sequence Seabrook5 Millstone-3 7 RSS13* RSSMAP15 52V 1D07 M-4 PWR-2 V-Sequence Xe 1.0 9.7(-1) 9.0(-1) 1.0 '1.0 Ol+I 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 2.4(-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 (Stu/hr) 140E6 4E6 70E6 20E6 0.5E6

• The same as M1A release category in Millstone-3.7 .

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The above principles were used to detemine source tem multipliers which would give a range of fission product leakage to the environment. A probabil-ity is associated with each source tem, and fo r later overpressurization f ailure modes (D, 379, and 577) the folicwing discrete probability distribu-tion is used -1.e.,

Subcategory Probability U-a .02 U-b .08 U-c .30 U-d .60 This indicates, for example that there is an 8% confidence level that U-b correctly defines the source tem for the U release category.

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

Particulate Eelease Factor (multiplier)

Probability U S3V S2V

.02 .22 .63 .17

.08 .071 .22 .07

.30 .024 .065

.02

.60 .0071 .021 .007 From this table it is seen that for the most likely release, i.e., "d", the reduction factors of the source tem are substantial.

The first two releases can be compared to releases published in BMI-2104 Volume V (Surry) for the TMLB'-c and AB-c sequences. These two sequences correspond to late contaiment failures and are both binned into 53 and S3vsequences. A comparison of these sequences is shown on Table 4.4 From this table it is evident that for the volatile species, Xe, Cs, and I, the release categories U and S3V bracket or exceed the mechanistic estimates carried out in BMI-2104 for both TM:B' and AB sequences. However, for the less volatile species Te, Ba, Ru, and La, the release of the AB sequence is the only one bracketed or superseded by the U and S3V releases. The release f ractio'n determined for the TMLB' sequence is higher than all the ST and S3V releases.

This discrepancy is primarily due to the comparatively early failure time. It is felt that agglomeration and settling would reduce the source for the TMLB' ~

sequence to values close to those reported for 53 and 53V. No comparative sequence for S2V was analyzed in BMI-2104.

In the case of the S6V release category a different probability distribu.

tion was used. This change reflects the break location, which initiates the s

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Table 4.4 Comparison of AB-c and TMLB'-c (BMI-2104) to Uf and U Release Fractions Release Probability Release Category Time (hrs) Xe Cs I Te Ba Ru La 53V-a .02 28 1.0 1.5(-2) 1.5(-2) 1.9(-2) 1.6(-3) 1.5(-3) 2.5(-4) 53V-b .08 36 9.0(-1) 5.3(-3) 5.3(-3) 6.6(-3) 5.7(-4) 5.1(-4) 8.6(-5)

S3V-c .30 54 8.0(-1) 1.6(-3) 1.6(-3) 2.0(-3) 1.7(-4) 1.5(-4) 2.5(-5) .

S3v-d .60 89 7.0(-1) 5.0(-4) 5.0(-4) 6.3(-4) 5.5(-5) 4.8(-5) 8.2(-6)

Y U-a .02 22 1.0 2.6(-2) 2.6(-2) 4.9(-3) 3.3(-3) 9.7(-4) 9.7(-5)

U-b ,.08 28 9.0(-1) 8.5(-3) 8.5(-3) 1.6(-3) 1.1(-3) 3.1(-4) 3.1(-5)

U-c .30 34 8.0(-1) 2.9(-3) 2.9(-3) 5.3(-4) 3.6(-4) 1.1(-4) 1.1(-5)

U-d .60 53 7.0(-1) 8.5(-4) 8.5(-4) 1.6(-4) 1.1(-4) 3.1(-5) 3.1(-6)

TMLB'-c -

12 1.0 2.8(-3) - 6.0(-4) 8.5(-2) 1.7(-2) 2.4(-5) 4.3(-4)

AB-c 24 1.0 4.8(-5) 4.7(-5) 4.0(-5)

, , 4.9(-5) 2.4(-7) 3.6(-5) t .

V-sequence. This break could be either in the hot-leg (b release subcategory) or the cold-leg (c release sub, category). This sequence is modeled as multi-puff release and each puff is treated separately. In this comparison only the su, of the release will be considered, since no adequate method of analyzing a multi-puff release is readily available. Table 4.5 shows a comparison be-tween the totals of tre various 00T releases and two Y-sequence releases com-puted for Surry and published in BMI-210c. One of the V-sequences is " dry,"

implying no water in the path of the release and the other is " wet," implying that the release passes through 3 feet of water before entering the atmo-sphere. From this comparison it can be seen that all the releases, except Cs l for the " dry" V-sequence, are bracketed by the 10f7 releases.

4.3 Recommended Source Terms The severe accident source terms used in the Seabrook Probabilistic Safe-ty Study reviewed in the previous sections, are aimed at the multi-puf f cr.n-sequence model present in the CRACIT computer code. In order to make these source terms useful to the NRC staff for evaluation with the CRAC code, total releases.must be used as summarized in Table 4.6. Furthermore, the suggested source terms of Table 4.6 together with their release category characteristics given in Table 4.7 have been adjusted to more closely represent our assessment / gg of the severe ac-idents based upon the RSS methodology.

It must also be noted that the suggested source term for the Steam Gener-

/4' ator Tube Rupture (SGTR) sequence is assumed to be one-tenth of the source term for the event V (56V). This is believed to be a conservative estimate and can be used in the absence of a more specific mechanistic calculation.

.. The suggested source terms can be used to estimate the number of health and ecor.omic effects (consequences) in.the population surrounding the Seabrook Station due to radioactive atmospheric releases as a result of core melt acci-

, dents. ,

The resulting consequences together with the frequency of radiological releases will enable the establishment of the severe accident risk at the Sea-brook site considering the double-reactor unit effect.

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Table 4.5 Comparison of 56V (sum) to V-sequence (Surry)

Release Fractions Release Probability Category ,Xe Cs I Te Ba Ru La S6v-a .02 .97 4.3(-1) 4.3(-1) 4.06(-1) 4.2(-2) 3.32(-2) 5.3(-3)

S6v-b .45 .97 2.95(-1) 2.95(-1) 1.36(-1) 3.53(-2) 1.52(-2) 2.0(-3)

$6V-c .45 .97 1.295(-1) 1.295(-1) 3.2(-2) 1.593(-2) 5.2(-3) 5.3(-4) e 1

S6V-d .08 - .97 '

5.2(-2) 5.2(-2) 1.3(-2) 6.6(-3) 2.0(-3) 2.2(-4) i

? -

l

) V (dry) ,

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

• I V -

1.0 1.04(-1) 3.84(-2) 2.5(-2) * * *

(submerged) r l'

• Individually not reported.

i .

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Table 4.6 BNL-Suggested Source Terms Release Category Xe O! I-2* Cs Te Ba Ru La 51 0.94 -

0.023 0.023 0.24 0.0033 0.41 9.8E-5 S2 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-8 3.8E-9 3.8E-10 SS 0.0091 -

3.5E-8 3.5E-8 6.1E-9 4.0E-9 1.2E-9 1.2E-10 56 0.90 -

3.6E-3 3.6E-3 6.7E-4 4.4E-4 1.3E-4 1.3E-5 3T 0.94 -

0.75 0.75 0.39 0.093 0.46 2.8E-3 ,

W 0.90. -

0.31 0.31 0.057 0.038 0.011 .1.1E-3 52v 1.0 -

0.31 0.31 0.32 0.034 0.025 4.2E-3 U 0.90 -

0.12 0.12 0.022 0.015 4.4E-3 4.4E-4 3Ti 1.0 -

0.024 0.024 0.030 2.6E-3 2.3E-3 3.9E-4 S4v 1.0 -

0.058 0.058 0.072 6.2E-3 5.4E-3 9.1E-4 4

55 0.014 7E-4 5.2E-7 5.2E-7 9.5E-8 6.3E-8 1.9E-8 1.9E-9 l .

56v 0.97 -

0.43 0.43 0.40 0.048 0.033 5.3E-3

$6v-d 0.90 -

0.043 0.043 0.040 4.8E-3 3.3E-3 5.3E-4

• Elemental iodine not used, all iodine treated as Cs!.

i

• • S6V-d release is 1/10th of the 56v values. , ,.

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Table 4.7 BNL-Suggested Release Characteristics for Seabrook Release Release Release Release Warning

• Release Time Duration Energy Height Time Category (hr) (hr) (Btu /hr) (m) (hr)

S1 1.9 0.5 140E6 10 0.35 52 2.6 1.0 0.5E6 10 1.05 53 66.0 0.5 250E6 10 63 SS 1.9 10 n/a 10 0.35

$6 4.5 4 0.5E6 10 0.50 10I 1.4 0.5 520E6 10 0.30 13i 27 10 10E6 10 26 52v 35 10 25E6 10 35 lET 27 0.5 250E6 10 26 S3v 81 0.5 450E6 10 76 S4v 50 0.5 250E6 0 -

49 ib 4.3 24 10E6' 10 0.30 S6v t.5 1.0 0.5E6 10 1.0

~~

56V-d 2.5 1.0 0.5E6 10 -1.0 i

l

• Warning time is defined as the time after core melt starts to the

! time of radiological release.

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, .'.f- .

5.

SUMMARY

AND CONCLUSIONS The purpose of this report is to describe the technical review of the Seabrook Station Probabilistic Safety Assessment and to present an assessment of containment perfomance, and radiological source term estimates for severe core melt accidents.

The contairment response to severe accidents is judged to be an important f actor in mitigating the severe accident risk. There is negligible probabil-Failure during the ity of prompt contaiment failure or f ailure to isolate. Most core melt accidents fi rst few hours af ter core melt is also unlikelyf would be effectively mitigated by contaiment spr y operation.

Our assessment. of the contaiment failure c)1aracteristics indicate that, there is indeed a tendency to fail containment through a realistic benign mode compared with the traditional gross failures.

The point-estimate release fractions usedlin the SSPSA are comparable in magnitude to those used in the RSS. In those' cases where comparisons can be made to the more mechanistic source tem study carried out by the Accident Source Tem Program Of fice (ASTP0) and reported in BMI-2104 it was found that -

for the most part similar to the the SSPSA releases were either higher than or[d that the energy of release was recent release fractions. It was also fou somewhat higher in the SSPSA than for other existing studies.

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" 'VJNN .

C omprican to >J AS H-/WO Se%< T)~yn W Ass - H o c(M) 6SPSA Tele sc Ty,<. "/. y c Iv1 '7., y ( tv.

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, REFERENCES Commonwealth Edison Company (September 1.

" Zion Probabilistic Safety Study," Electric Co.

Study," Philadelphia 1981). Safety f Probabilistic

2. " Limerick (September 1982). Study," Power Authority of the State

" Indian Point Probabilistic i Safetyn Company (March Utilities 1982).

3. Study," Northeast of New York and Consolidated Probabilistic Safety Ed so

" Millstone Unit 3 Safety 4.

(August 1983). Station Probabilistic et al . , "Se abrook Inc., PLG-0300 (December 1983).

ik B. J.. Ga rrick ,

5. Assessment," Pickard, Lowe and Garr c ,the Seabrook Station Probabi 6.

A. A. Garcia, et al . , "A Review ofLawrence Livermore National La Safety Assessment,"

12,1984). d Evaluation ofRadiological the Millstone Unit Failure Modes,(report to be

7. _ M. Khatib-Rahbar, Safety Study: et Containments al ., " Review an

" NUREG/CR , 4143 ProbabilisticSource Tems' and Of f-Site Consequence ,

published). Heltdown Accident Response C

8. R. O. Wooten and H. Avei, " MARCH:istics Code D An Improved Model for Molten (1980).

Muis, et al . , "CORCON-Modi:2415 (1981).

J. F.

Core / Concrete Interactions," SAND 80-Estimation of Pre-Exist-9 E. Hall, " An and Vent Valve Leakage Areas Re-B. E. Miller, A. K. Agrawal, and "R. A-3741,11/15/84 (Draft report 29, 1984.

10.

ing Cantainment Leakage Areas and Purgesulti i

i dated August 1984) transmitted v a f.oundary He land (May 14,1984).

11. W. Lyon (organizer), "RCS Pressuredents," US

" Estimates of ~arly ContainmentREG-1079 (DraftWASH-1400, 1985).

12.

taiment loads Working Group, NUU.S. Nuclear Regulat

" Reactor Safety Study ," 75). ,

and Indian" 13.

i NUREG-75/014 (October of Core 19 Melt Accidents at the Zionies for Mi

" Preliminary Assessment

14. Point Nuclear Power Plants 1). and Strateg Application NUREG-0850, Vol. I _ (November Safety 198 Study Methodology al., " Reactor 59/2 of 4.

G. S. Kolb, et Oconee #3 PWR Plant ," NUREG/CR-16 15.

Program:

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