ML20132A832
| ML20132A832 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 09/19/1985 |
| From: | Knighton G Office of Nuclear Reactor Regulation |
| To: | Harrison R PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| References | |
| NUDOCS 8509260027 | |
| Download: ML20132A832 (70) | |
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Docket Nos.: 50-443 SEP 19 985 DISTRIBUTION and 50-444 . Docket: En e 50 545/444 E F ?
'NRC PDR RPerlis, ELD LPDR EJordan NSIC JPartlow Mr. Robert J. Harrison PRC System BGrimes President & Chief Executive Officer LB#3 Reading WLyons Public Service Company of New Hampshire JLee JRosenthal Post Office Box 330 VNerses BSheron Manchester, New Hampshire 03105 HLThompson/FMiraglia TMNovak
Dear Mr. Harrison:
GWKnighton
Subject:
Seabrcok Probabilistic Safety Assessment (PSA) Review In my letter of April 4,1985 to you, I provided you a status at that time of our PSA review effort and our decision to terminate this review. As an enclosure to that letter, I provided you a draft review report from one of our contractors, Lawrence Livermore Laboratories (LLNL). The LLNL draft report constituted the Phase I review (review of the sequences leading up to core melt) of the Seabrook PSA. As noted i' n my letter, because your support of our review had to be withdrawn, it impeded our efforts to complete the review. In your letter of July 12, 1985, you notified us of your recent contract with your PSA consultants, and you were prepared again to provide our staff any necessary support of their PSA review. Subsequently, we completed the Phase II review (review of the severe accident response and consequences portion of.the PSA), and for your information, enclosed is a copy of the draft report from our Phase II contractor, Brookhaven National Laboratory. We intend to meet with your staff to discuss our Phase II review 'and this draft report. Sincerely,
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hDR $ 43 George W. Knighton, Chief A PDR Licensing Branch No. 3 Division of Licensing ,
Enclosure:
As stated cc: See next page f
/yt ton 9//f/85 9/ /85
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- l Mr. Robert J. Harrison . Public Service Company of New Hampshire Seabrook Nuclear Power Station cc:
Thomas Dignan, Esq. E. Tupper Kinder, Esq. John A. Ritscher, Esq. G. Dana Bisbee, Esq. Ropes and Gray Assistant Attorney General 225 Franklin Street Office of Attorney General Boston, Massachusetts 02110 208 State Hosue Annex Concord, New Hampshire 03301 Mr. Bruce B. Beckley, Project Manager Public Service Company of New Hampshire Resident Inspector Post Office Box 330 Seabrook Nuclear Power Station Manchester, New Hampshire 03105 c/o US Nuclear Regulatory Comiss. ion Post Office Box 700 Dr. Mauray Tye, President Seabrook, New Hampshire 03874 Sun Valley Association 209 Sumer Street Mr. John DeVincentis, Director Haverhill, Massachusetts 01839 Engineering and Licensing Yankee Atomic Electric Company Robert A. Backus, Esq. 1671 Worchester Road 0'Neil, Backus and Spielman Framingham, Massachusetts 01701 116 Lowell Street Manchester, New Hampshire 03105 Mr. A. M. Ebner, Project Manager United Engineers & Constructors Ms. Beverly A. Hollingworth 30 South 17th Street 7 A Street Post Office Box 8223 Hampton Beach, New Hampshire 03842 Philadelphia, Pennsylvania 19101 William S. Jordan, III Mr. Philip Ahrens, Esq. Diane Curran Assistant Attorney General Harmon, Weiss & Jordan State House, Station #6 20001 S Street, NW Augusta, Maine 04333 Suite 430 Washington, DC 20009 Mr. Warren Hall Jo Ann Shotwell, Esq. Public Service Company of Office of the Assistant Attorney General New Hampshire Environmental Protection Division Post Office Box 330 One Ashburton Place Seabrook, New Hampshire 03874 Boston, Massachusetts 02108 Seacoast Anti-Pollution League D. Pierre G. Cameron, Jr., Esq. Ms. Jane Doughty General Counsel 5 Market Street Public Service Company of New Hampshire Portsmouth, New Hampshire 03801 Post Office Box 330 Manchester, New Hampshire 03105 Mr.. Diana P. Randall 70 Collins Street Regional Administrator, Region I Seabrook, New Hampshire 03874 U.S. Nuclear Regulatory Comission 631 Park Avenue King of Prussia, Pennsylvania 19406
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8 Public Service Company of Seabrook Nuclear Power Station New Hampshire cc: Mr. Calvin A. Canney, City Manager Mr. Alfred V. Sargent, City Hall Chairman 126 Daniel Street Board of Selectmen Portsmouth, New Hampshire 03801 Town of Salisbury, MA 01950 Ms. Letty Hett Senator Gordon J. Humphrey Town of Brentwood U. S. Senate RFD Dalton Road Washington, DC 20510 Bre7twood, New Hampshire 03833 (Attn: Tom Burack) Ms. Roberta C. Pevear Senator Gordan J. Humphrey Town of Hampton Falls, New Hampshire 1 Pillsbury Street Drinkwater Road Concord, New Hampshire 03301 Hampton Falls, New Hampshire 03844 (Attn: HerbBoynton)
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Ms'. Sandra Gavutis Mr. Owen B. Durgin, Chainnan
._ Town of Kensington, New Hampshire Durham Board of Selectmen RDF 1 Town of Durham East Kingston, New Hampshire 03827 Durham, New Hampshire 03824 Charles Cross, Esq.
Chainnan, Board of Selectmen Shaines, Mardrigan and Town Hall McEaschern South Hampton, New Hampshire 03827 25 Maplewood Avenue Post Office Box 366 Mr. Angie Machiros, Chainnan Portsmouth, NH 03801
. Board of Selectmen for the Town of Newbury Newbury, Massachusetts 01950 Mr. Guy Chichester, Chaiman Rye Nuclear Intervention Ms. Cashman, Chairman ~ Comittee Board of Selectmen c/o Rye Town Hall Town of Amesbury 10 Central Road Town Hall Rye, New Hampshire 03870 Amesbury, Massachusetts 01913 l Jane Spector Honorable Richard E. Sullivan Federal Energy Regulatory Mayor, City of Newburyport Comission Office of the Mayor 825 North Capital Street, NE City Hall Room 8105 Newburyport, Massachusetts 01950 Washington, D. C. 20426 Mr.. Donald E. Chick, Town Manager Mr. R.- Sweeney Town of Exeter New Hampshire Yankee Division 10 Front Street Public Service of New Hampshire Exeter, New Hampshire 03823 Company 7910 Woodmont Avenue Mr. William B. Derrickson Bethesda, Maryland 20814 Senior Vice President Public Service Company of New Hampshire- ,
i Post Office Box 700, Route 1 ! Seabrook, New Hampshire 03874
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NUREG/CR-
- w h j i A REVIEW 0F THE SEABROOK STATION PROBABILISTIC SAFETY ASSESSMENT: CONTAINMENT FAILURE MODES AND RADIOLOGICAL SOURCE TERMS M. Khatib-Rahbar, A. K. Agrawal, H. Ludewig and W. T. Pratt September 1985 Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973
111 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 unlikely. (4) the point-estimate radiological releases are comparable in magnitude to those used in WASH-1400, and (5) the energy of release is somewhat higher than for the previously reviewed studies. ere I ) I
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l $ iv ACKNOWLEDGMENT
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-v- l CONTENTS Page iii ABSTRACT................................................................
ACKNOWLEDGMENT.......................................................... iv LIST OF TABLES.......................................................... vi vii LIST OF FIGURES.........................................................
- 1. INTRODUCTION.......................................................
I I 1.1 Background.................................................... 1 1.2 Obj e c t i v e s a n d Sc o pe . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.3 Org ani za ti on of the Repo rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
- 2. PLANT DESIGN AND FEATURES IMPORTANT TO SEVERE ACCIDENT ANALYSIS....
2.1 As s e s sm e n t o f Pl a n t De s i g n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.2 Compa ri s o n wi t h Ot he r Pl a nt s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
- 3. ASSESSMENT OF CONTAINMENT PERFORMANCE.............................. 7 7
3.1 Background.................................................... 3.2 Co nt a i nme nt F a i l u re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.2.1 Background............................................. .8 3.2.2 De s i gn De s c r i p t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.2.3 Leakage Rate Calculation............................... 21 3.2.4 Cont a i nme n t Fa il u re Mod el . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.2.4.1 Le a k-Be fo re-Fa il ure. . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.2.4.2 Cl a ssi fi cation of Fail ure. . . . . . . . . . . . . . . . . . . . . 23 3.2.5 Co ntai nment Pre s 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 Contai nment Fail ure Probabil ity. . . . . . . . . . . . . . . 31 3.2.5.5 Co n t a i nme nt En cl o s u re . . . . . . . . . . . . . . . . . . . . . . . . . 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 Cont a i nme nt Ma t ri x (C-Ma t ri x ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 3.6 Rel ease Catego ry Frequenci es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44
- 4. ACCIDENT SOURCE TERMS.............................................. 48 4.1 Assessment of Severe Accident Source Te rms. . . . . . . . . . . . . . . . . . . . 48 4.2 Source Te rm Uncertai nty Analysi s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 4.3 Re c omme n d ed So u rc e Te rm s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56
- 5.
SUMMARY
AND CONCLU S ION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 60
- 6. REFERENCES......................................................... 61 l
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-vi-LIST OF TABLES Table Title Page 2.1 Comparison of Selected Design Characteri stics.. . .. .. . .... .. .... . . 5 3.1 Containment Operating and De sign Parameters. . . . . . . . . . . . . . . . . . . . . . 10 3.2 Co nt ai nment Li ne r Penet rati on s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 3.3 Leak Area Estimates fo Mechanical Penetrations...... . ... . ... ... .. 29 3.4 Frequencies of Occurrence of the Plant Damage States............. 35 3.5 Contai nment Re sponse Cl ass De fi ni tions . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 3.6 Con tai nment Cl a s s Mean Frequenci e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 3.7 Accident Phase 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 Seab rook. . . . . . . . . . . . . . . . . . . . . . . 41 3.10 Frequency of Dominant Release Categories (yr-1).... . . ..... . .. ... . 45 3.11 Contribution of Containment Response Classes to the Total Co r e Mel t F r e q u e n cy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 3.12 Release Category Frequency as a Fraction of Core Melt Frequency........................................................ 47 4.1 . Seabrook Poi nt-Estimate Rel ease Categori es. . . . . . . . . . . . . . . . . . . . . . . 49 4.2 Late Ov erpressuri zation Fail ure Compa ri son. . . . . . . . . . . . . . . . . . . . . . . 51 4.3 Comparison of Releases for Failure to Isolate Containment a nd t he By- Pa s s Seque nc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 4.4 Compa r i son o f AB- c a nd TMLB ' - c (BM I-2104 ) to 1"PT a nd 37. . . . . . . . . . 55 4.5 Comparison of 337 (sum) to V-sequence (Surry).................... 57 4.6 B NL-Sug g e sted So u rce Te rm. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 4.7 BNL-Suggested Release Characteri stics for Seabrook. . . . . . . . . . . . . . . 59
-vii-LIST OF FIGURES Figure Title Page 3.1 A schematic representation of source term calculation............ 9 Equi pment hatch wi th personnel ai rl ock. . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.2 14 3.3 Pe r s o n n el a i rl o c k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Typi cal high energy pi pi ng penetration. . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.4 Typical moderate ene rgy pi pi ng penetrati on. . . . . . . . . . . . . . . . . . . . . . . 16 3.5 Typi cal el ectri cal penetrati on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.6 3.7 Typical ventil ation penetrati on. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.8 A pictorial representation of leakage categories................. 25 3.9 Estimated radial displacement of containment wall . .. . .... ...... . . 26 3.10 Estimated containment failure fractions.......................... 32 3.11 Definitions 'of the plant damage states used in SPSS..........;.... 34 e 4 Oe d Fe
- 1. INTRODUCTION
1.1 Background
Probabilistic Risk Asessment (PRA) studies have been undertaken by a num-ber of utilities (as exemplified by Refs.1-4) and submitted to the Nuclear Regulatory Commission (NRC) for review. Brookhaven National Laboratory (BNL) under contract to the NRC, has .been involved in reviewing core melt 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 tne Public Service Company of New Hampshire and Yankee Atomic Electric Company in December 1983.5 1.2 Objective and Scope The objective of this report is to provide a pers'pective on severe acci-dent propagation, contaiment response and failure modes together with radiol-ogical source tem 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.
? i n the pres report, pri ncipal containment design features are dis-cussed and compared 'th + hose of Zion, Indian Point and Millstone-3 designs.
Those portions of th PSS Telated to severe accident phenomena, containment response and radiologic source tems are described and evaluated. Numerical adjustments to the SPSS estimates are documented and justified. 1.3 Organization of the Report At brief review of the Seabrook plant features important to severe acci-dent analysis is presented in Chapter 2 along with comparisons to Zion, Indian Point and Millstone-3 plant designs. Chapter 3 contains the assessment of containment pe rfo rmance. Specifically, definition of contaiment response classes and plant damage states, analytical ~eho containment failure model, containment event tree and accident phenomeno y and the containment matrix are reviewed. Chapter 4 addresses the accide t source tems together with justifications for adjustment where necessary. The results of this review are sumarized in Chapter 5.
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- 2. PLANT DESIGN AND FEATURES IMPORTANT TO SEVERE ACCIDENT ANALYSIS In this section, those plant design features that may be important to an assessment of degraded and core melt scenarios and containment analysis are reviewed. These important features are then compared with the Zion, Indian Point and Millstone-3 facilities to identify commonalities for benchmark compa riso ns.
2.1 Assessment of Plant Design The Seabrook Station is comprised of two nuclear units each having an identical Nuclear Steam Supply System (NSSS) and turbine generator. The units are arranged using a " sling-along" concept which results in Unit 2 being , . arranged similar to Unit 1 but moved some 500 feet west. Each unit is a 1150 MWe (3650 MWt)', 4-loop, Westinghouse PWR plant. The turbine-generators are supplied by the General Electric Company and the balance of the plant is designed by United Engineers and Constructors. t Each containment completely encloses an NSSS, and is a seismic Category I , reinforced concrete structure in the form of 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 appror; mate dimensions of the containment are:
Inside diameter 140 ft.
'Inside height 219 ft.
Vertical wall thickness 4 ft. 6 in. and 4 ft. 7 1/2 in. Dome' thickness 3 f t . 6 1/8 i n. Foundation mat thickness 10 ft. Containment penetration; are provided in the lower portion of the structure, and consist of a personnel lock and an equipment hatch / personnel lock, a fuel transfer tube, electrical, instrumentation, and ventilation penetrations. Each contairrnent enclosure (also known as secondary containment) su r-rounds a containment and is designed in a similar configuration as a vertical right cylindrical seismic Category I, reinforced concrete structure with dome and ring base. The approximate dimensions of the structure are: inside diam-
- eter,158 ft; vertical wall thickness, varies from 1 ft, 3 in. to 3 ft; and
! dome thickness,1 ft, 3 in. a The containment enclosure is designed to collec any leakage from the
/) containment structure other than leakage associated th piping, electrical
( and access passage penetration and discharge to the filtration system of l r contai nment. To accomplish this, the space between the containment 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 containment spray system is utilized for post accident containment heat removal. 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 containment dome. After a prescribed amount 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 generated at a designated containment pressure. The system is completely re-dundant and is designed to withstand any single failure. The containnent isolation system establishes and/or maintains isolation of the containment 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 adequate shutdown mar-gin. The ECCS also provides continuous long-term post-accident cooling of the core by recirculating borated water between the containment sump and the reat-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. The control building contains the building 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 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 perfonn a safety-related function following a design basis accident are discussed below:
- a. The containment enclosure exhaust filter system for each unit col-lects, filters and discharges any containment leakage. The system is not normally in operation, but in the event of an accident, it is placed in operation and- keeps the containment enclosure and the building volumes associated with the penetration tunnel and the ESF equipment cubicles under negative pressure to ensure all leakage from.
the containment structure is collected and filtered before discharge to the plant vent. e
- 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 f rom the con-densate water storage tank to the four steam generators upon loss of normal feedwater flow to remove heat from the reactor coolant system. Operation of the system will continue until the reactor coolant system pressure is. reduced to 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 sources 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 states. The normal preferred source of power for each unit is its own main turbine generator. The redundant safety feature buses of each unit are power-ed by two unit auxiliary transformers. A highly reliable generator breaker is provided to isolate the generator from the unit auxiliary transformers 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 powered by two reserve auxiliary transformers. Upon loss of off-site power, each unit is supplied with adequate power by either of two fast-starting, diesel-engine generators. Either diesel-engine generator and its associated safety feature bus is capable of providing adequate 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-ed through the redundant 125 volt direct current buses and their associated battery banks, battery chargers and inverters. 2.2 Comparison 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 seen that the containment 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 nearly iden-tical between the four units. The Seabrook contai nment building basemat and the internal concrete structures are composed of basaltic-based concrete. As concrete is heated, water vapor and other gases are released. The initial gas consists largely of carbon dioxide, the quantity of which depends on the amount of calcium carbon-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
. -5.)
Table 2.1 Comparison of Selected Design Characteristics Zion Indian Point Millstone Seabrook Design Parameters Unit 11 Unit 2 3 Unit 3 .7 Unit 1,2 6 R: actor Power [MW(t)] 3,250 3,030 3,411 3,650 Containment Building: Frae Volume (ft ) 3 2.73 x 10 6 2.61 x 10 6 2.3 x 10 6 2.7 x 10 6 Design Pressure (psia) 62 62 59.7 67.7 Initial Pressure (psia) 15 14.7 12.7/9.1 15.2 Initial Temperature (*F) 120 120 120/80 120 Primary System: 3 12,710 11,347 11,671 13,14'O i Water Volume (ft,) Steam Volume (ft ) 720 720 ? 2,012 222,739 Mass of UO 2 in Core (1b) 216,600 216,600 222,739 Mass of Steel in Core (1b) 21,000 20,407 ? 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 Head 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 (lb) 200,000 173,000 348,000 213,000 Initial Pressure (psia) 665 665 600 615 Temperature ('F) 150 150 80 ? Refueling Water Storage Tank: 7 Total Mass of Water (ib) 2.89 x 10 6 2.89 x 10 6 10 2.89 x 106
- T;mperature ('F) 100 120 50 86 Rsactor Cavity:
Configuration Wet Wet Dry Dry / Wet i Ccncrete Material Limestone Basaltic Basaltic Basaltic 6
- Minimum (Maximum Capacity = 3.9 x 10 lb)
result of corium/ concrete interactions would be expected to take a very long time.
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- 3. ASSESSMENT OF CONTAINMENT PERFORMANCE 4
In'this chapter, the review of containment responce to severe accidents l is described. Analytical techniques used to analyze core ~ meltdown phenomena and containment response are reviewed, containment failure model is assessed and plant damage states and containment f ailure modes are evaluated. Parallels between this study and other PRAs are set forth. Finally, the rel-evance and validity of the conclusions is addressed. 3.1 Contaiment Analysis Methods , 4 A brief description of the computer codes used to perfom the transient degraded, core meltdown and containment response analyses is provided in this : section. ! 4- The MARCHs computer code is used to model the core. and primary system '
- transient behavior and to obtain mass and energy releases from the primary i system until reactor vessel failure. These mass and energy releases are then used as input to the 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 M00MESHS computer code is used. This code calculates the steam and hydrogen blowdown from the reactor vessel using an isothermal ideal gas model. The water level boil-off from the reactor cavity floor is modeled using a saturated critical heat flux correlation. Additionally, the accumulator discharge following de-pressurization caused by the vessel failure is also considered. A modified version of the CORCONS code is used to replace the INTER8 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 j proper sequencing and integration into the overall mass and energy input to ! C0C0 CLASS 95 code. C0C0 CLASS 9, a modified version of the Westinghouse C0C0 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 contaiment sprays and sump recirculation.
- Fission product transport and consequence calculations are - performed ;
using the CORRAL-II and the PLG proprietary CRACITS computer codes, respec-tively. The analytical methods used to carry out the core and contaiment thermal
- hydraulics, and fission product transport calculations are identical to those i used.for MPSS-3.7
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3.2 ' Containment Failure 3.2.1 Background In order to assess the risk of the Seabrook-1 plant, radiological source terms have to be calculated. Many steps are involved in such calculations. ; These are schematically shown in Fig. 3.1. The mode and time of containment , failure directly impact on the radioactivity release categories. These, when ' coupled with the status of reactor cavity and the spray system, determine the source terms. This section deals with the mode and time of containment fail-u re. i 3.2.2 Design Description The primary containment of the Seabrook plant is a seismic Category I re-inforced concrete' dry structure. It consists of an upright cylinder topped , 4 with a hemispherical dome. The inside diameter of the cylinder is 140 feet i and the inside height from the top of the basemat to the apex of the dome is approximately 219 feet. The cylindrical wall .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 L ai rlock. The net free volume of the containment is approximately 2.7 x 106 ft 3. (
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The inside of the containment is welded with a. steel liner. The liner plate is the cylinder is 3/8" thick in all areas except penetration and the ( tion of the .basemat and cylinder where it is 3/4" thick. This liner , e serves as a leak-tight membrane. Welds that are embedded in the concrete and ! not readily accessible are covered by a leak' chase system which permits leak j testing of these welds throughout the life of the plant. The dome liner is
- 1/2" thick and flush with the outside face of the cylindrical liner. The i operating and the design parameters of containment are noted in Tabl'e 3.1.
The containment building is surrounded by an enclosure. The containment enclosure is a reinforced concrete cylindrical structure with a hemispherical The inside diameter of the cylinder is 158 feet. The vertical wall dome. varies in thickness from 36 inches to 15 -inches; the dome is 15 inches thick. The inside of the dome is 5'6" 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. i The containment 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 negative pressure (-0.25" water gauge) during accident conditions by fans which take suction from -the containment- 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:
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4 CONTAINMENT TIME OF FAILURE FAILURE MODE WET OR DRY RELEASE SPRAY j REACTOR CAVITY CATEGORY SYSTEM SOURCE TERM J 4 l Figure 3.1 A schematic representation of source term calculation.
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Table 3.1 Containment Operating and Design Parameters Parameter Value Normal Operation Pressure , psig 0.5 Inside Temperature , F 120 Outside Temperature , F 90 Relative Humidity , % 45
- Service Water Temperature , F 80 Refueling Water Temperature F 86 Spray Water Temperature , F 88 Containment Enclosure Pressure , inches w.g. -0.25 Design Conditions l Pressure , psig 52.0 Temperature , F 296 Free Volume , ft3 2.7x106 Leak Rate , % mass / day 0.2 Containment Enclosure Pressure , psig ' -3.6 l
l 4
_11
- 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, anchor piping, or furnish some other operational 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 (inches) Barrel 3 1/2 Spherical 1 3/8 Flange 5 3/8 Air lock mounting 1 1/2 sleeve The equipment hatch cover is fitted .with two seals that enclose a space which can be pressurized to 52.0 psig. The flange of the cover plate is at-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. Significant dimensions of the air lock are as follows: Pa rameter 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 test 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 !
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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. Personnel Air Lock The personnel air lock (Fig. 3.3) consists of the air lock doors (2) and the lock barrel. The barrel, which is also the sleeve for the personnel air lock, is imbedded in the shell of the concrete containment. The centerline of the barrel is located at elevation 29'6" and an azimuth of 315*. Significant dimensions are as follows: Parameter Dimensions Clear Opening 7'0" 0.D. of Flange on Door 7' 9 1/8" 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 containment. 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 ri ng , breach-type mechanism. These doors are also mechanically interlocked so that only one door can be opened at a time. The capability exists for bypassing 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 ene 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. High energy piping penetrations (Fig. 3.4) consist of a section of pro-i 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 while permitting relative motion between the pipe and the support. The annu-lar space between the process pipe and the sleeve is completely filled with fiberglass thermal insulation. The pipe and the fluid head, are classified as ASME III Safety Class 2 (NC), whereas the sleeve is classified as part of the concrete containment, 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 containment penetration sleeve, and a flat circular end-plate. The pipe is classified as ASME III Safety Class 2 (NC). The sleeve is classified as ASME III Div. 2 (CC). The end-plate is classified as Class MC. 1
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_17 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 penetrations (Fig. 3.6) consist of a stainless steel header plate with an attached tenninal box, electrical modules which are clamped to the header plate, and a carbon steel weld ring which 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 Class MC components (NE); that portion of the sleeve which is backed by concrete was designed as part of the concrete containment, ASME III (CC). Double silicone and Hypalon 0-rings provide a seal with a cavity fo r leakage monitoring between the header plate and the modules. The header plate is provided with a hole on the outside of the containment 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 fluid. 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 l The fuel transfer tube assembly consists of the fuel transfer tube, the penetration sleeve, the fixed saddle on the reactor side, and tne 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 about 25". Ventilation Penetrations i There are two types of ventilation penetrations -- the containment air purge penetrations (HVAC-1 and HVAC-2) and the containment on-line penetra-tions (X-16 and X-18). The contairinent 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 end with 36" weld neck flanges and, attached to these flanges, the inner and outer isolation valves. Together with the pipe, these valves fonn a part of the containment pressure boundary. The valves are 36" diameter butterfly valves with fall-safe pneu-matic operators. The weld between the pipe and the contairinent liner is equipped with a leak chase for pressure testing. 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 contairinent, and a 3/4" valve and test connection is attached to it. The
.. o ,' Table 3.2 Containment Liner Penetrations 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 injection 8" 4
X-14 to X-15 Containment building spray 8" X-16, X-18 Containment on-line purge 8" X-17 Hydrogenated vent header 2" X-20 to X-23 CCW supply and return 12" X-24 to X-27 Safety injection 4" X-28 to X-31 CVCS to pump seal injection 2" X-32, X-34 Drain line 3",2" X-33, X-37 CVCS 3" X-35, X-36, X-40 RCS test / sample control 1" or smaller X-52, X-71, X-72 X-38 Combustible gas control 10" X-39 Spent fuel pool cooling 2" X-43, X-47, X-50 Instrumentation lines ? X-57 X-60, X-61 From containment recirculation sump 16" X-62 Euel transfer tube 20" , X-63 to X-66 Steam generator blowdown 3" X-67 Service air ' 2" HVAC-1,2 Containment purge supply / exhaust lines 36" X-19, X-41, X-42 X-44 to X-46, X-48 ~ $ pare ? X-49, X-51, X-58 X-59, X-68 to X-70
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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 containment 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 Leakage Rate Calculation Under severe accident conditions the pressure inside the containment quickly builds Jp in the range of 75 to 200 psi. At these pressures, any leakage through the containment holes will essentially be choked. The leakage under choked flow condition is given as (Ref.10): k+1 W= k(k 1) Av7P (1) I 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 absolute temperature, Then Eq.(1) becomes k+1 W= k(k ) A (3) The mass of mixture can be written as M = Vp or, M=h (4) where V is the free mixture volume in the containment. Equations (3) and (4) can be combined to get the leakage rate, in terms of mass fraction, as
k+1 f= k(k 1) A (5) Note that the leakage rate, when expressed in tems of mass fraction, depends only on the leakage area. For Seabrook-1, using V = 2.704x10 6 ft 3 and T = 296 F, Eq.(5) gives Leakage ~ Rate = 0.721 Ai n W/o per hour (6) 2 where A i n is the leakage area in in . Al te rnately, Leakage Rate = 17.3 Ai n W/0 Per day. (7) The essentially intact design basis containment leakage of 0.2 w/o per . day, thus, corresponds to an equivalent leakage area of 0.012 in 2 (or, an equiva-lent hole of 1/8-in diameter). A leakage area of 4 to 10 in would correspond 2 to the leakage rate of 2.9 to 7.2 w/o per hour. In other words, it will take about 14 hours to leak the entire content to the environment through a 10-in 2 hole. 3.2.4 Containment Failure Model 3.2.4.1 Leak-Bef o re-F ailu re During accident sequences involving core damage, the containment struc-ture will be exposed to pressures and temperatures beyond those used in the design basis accident (DBA). Response of the containment 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 leakage from the containment to the surroundings. This mode of containment failure is temed local failure. The containment leakage can occur at many locations and discontinuities such as mechanical and electrical penetrations, personnel lock, equipment hatch, fuel transfer tube, welds, and in between the liner and concrete. Depending upon the size of leakage area and the accident sequence, local failures may gradually relieve pressure., thereby gross con-tainment failure may be averted. The leak-before-failure model is a realistic one. The extent of leakage and the health consequences must, however, be carefully studied. In order to explain this issue, it is observed that traditionally probabilistic risk as-sessment is made by using what is temed a threshold model. In the threshold model, the containment is considered intact until the internal loading equals or exceeds a pressure threshold (which may also be temperature dependent), at which it is deemed to have suffered a failure (gross). If the internal load-ing is below this threshold value, the containment is considered intact and hence the risk is quite low. In the leak-before-failure model, the release of activity, which is considerably small compared with that for the gross failure 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.
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4 3.2.4.2 Classification of Failure The SSPSA report has classified containment failures in three categories:
. Containment Failure Category A. Includes containment failures that 1
develop a small leak that is substantially larger than the leak ac-i ceptable from an intact containment, but not large enough to arrest the pressure rise in the containment. Category A failures thus cause an early increase in the rate of leakage of radionuclides over the de-l sign basis leak rate but pressurization of the contairunent continues j until either a category B or C contaiment failure occurs. l The intact containment is defined as the one in which leakage is lim-ited by the Technical Specification value. For Seabrook-1, this value
- is 0.2 w/o per day at the calculated peak accident pressure of approx-l imately 47 psig. Note that the SSPSA study has used 0.1 volume per-cent per day for this leakage, although prior to the most recent amendment dated August 1984, the FSAR has 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 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 approximately 1 to 2 hours. 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 tenninate the pressure rise before a category C containment failure occurs.
The definition of " substantial" fraction is unclear.
. Containment Failure Category C. Includes those contat rynent failure I
modes tnat develop a large leak area. A large fraction of the total
; radionuclide source tenn is released over a period of less than 1 hour. All gross failure modes are included in category C.
Mathematically, these three failure categories can be expressed in tenns j of leakage areas as follows: < 1 l [A 0BA < AA ANP Type A j AB < AP Type B (8)
, e Merv ^- AC < AP Type C where yg/ ' [ff ADBA = leakage area corresponding to the technical specification limit for containment leakage,
*There appears to be substantial update / changes in the -Engineered Safety
, Features flow diagram, including arrangements of motor operated valves and l bypass lines, which may substantially change the frequency of events. BNL,
- however, is not reviewing this part of SSPSA.
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( gaT g/ ANP = leakage area not large enough to arrest
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- t. The leakage area required to release a subst. al fraction of the radio-9 nuclide source tem in approximately an hour an can be computed using Eq.
(6). Assuming one-hundred percent turnover as a substantial fraction in one hour Eq. (6) gives the required leakage area to be equal to 138 in2 or about g\ a 1 ft . Therefore, any containment leak area in excess of21 f t will be N kO fined as a gross _ containment failure -(Cateaory C).- This estimate of the ead ' g
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a(factor two too high from the value stated inTPS The leakage area required to arrest containment pressurization is in the range of 4 to 10 square inches, the lower value being more representative of wet sequences and the upper value is representative of dry sequences. A leak area of about 6 square inches will result in the release of about 100 w/o of activity in a day (see Eq. 7). The upper bound leak area for Type A failure is taken as 4 in . This corresponds to release of the radioactive source term (100% turnover in about 36 hours. The Category B leak area is, thus, in the range of 4 in} to 1 ft2. Figure 3.8 is a pictorial representation of these leakage categories. 3.2.5 Containment Pressure Capacity 3.2.5.1 Concrete Containment The 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 containment. The containment structure includes the cylindrical wall, the hemispherical dome, the base slab and the base slab and containment wall junction. The most critical membrane tension was found to occur in the cylinder in the hoop direction. The median pressure which causes yield of both the liner steel and the reinforcing bars was found to be approx-imately 157 psi, with a coefficient of variation of 0.084. The ultimate hoop load in cylinder is 216 psig. The containment wall is, tnus, assumed to fail at this pressure. At pressures beyond this, very large irreversible defoma-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 4.7% (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, with a logarithmic standard deviation of 0.25. However, the shear mode of failure is more restrictive. For this mode, the median failure pressure is estimated in SSPSA as 323 psig, with a logarithmic standard deviation 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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27-an estimation of leak areas was, therefore, not considered for this mode of failure. Secondary stresses in the cylindrical portion of the containment occur at discontinuity such as at the base slab containment wall junction, at the springline, and where the amount of reinforcing changes. The flexural yield at the base of the cylinder occurs at 175 psi. At higher pressures, a plastic hinge forms with considerable cracking of the concrete. These cracks, how-ever, are small enough so as not to threaten the integrity of the liner. The
~
loss of integrity of the liner is not expected until a median pressure of 408 psi is reached. Thus, the failure of the base slab and containment wall junc-tion is not limiting. In summary, the containment wall is expected to undergo significant de-formation (=4.7% ar/r) prior to its failure at 216 psig. At this pressure, Type C (i.e., gross) failure occurs. 3.2.5.2 Liner The elongation capacity of the steel liner is computed by neglecting the f riction forces between the liner and the-concrete. The possibility that the liner stresses and strains could be different between two different pairs of tees was, however, considered. The SSPSA computed an elongation of 8.1 per-cent under uniaxial conditions, or an elongation of 4.7 percent under plane strain conditions can be achieved without fracture. This would ensure integ-rity of the liner until fracture of the reinforcing bars. Additionally, the leakage of the containment at penetrations is considered likely before hoop failure of the liner occurs. 3.2.5.3 Penetrations At all major penetrations, the containment wall is thickened and addi-tional reinforcement is provided to resist stress concentrations. None of the meridional or hoop reinforcing bars are terminated at penetrations. Instead, they are continued around the penetrations, thus ensuring that excess hoop and meridional capacity is available. Table 3.2 lists all piping penetrations. As the containment pressure increases beyond its yield value (157 psi), large radial deformations begin to occur. This induces stresses in the pipes by relative displacements between the containment wall and the pipe whip re-straints. Therefore, the most critical penetrations are the areas where the pipe is supported close to the penetration. Also, stronger and stiffer pipes develop higher forces at the penetrations for a given relative displacement. The SSPSA study selected the following penetrations for investigation as being among the lines most likely to fail: Penetration X-23 12" schedule 40 carbon steel (also X-20 to X-22 by similarity) Penetration X-26 4" schedule 160, stainless steel (al so X-24, X-25, X-27)
Penetration X-71 1" - multiple pipe penetration (also X-72 and possibly others) Penetration X-8 18" main feedwater schedule 100 (also X-5 to X-7) carbon steel Fuel Transfer Tube Convoluted Bellows The probability of failure at these penetrations was computed by (a) establishing a pressure-displacement relation, (b) estimating the failure probability as a function of radial displacement and.then (c) combining the two. The radial displacements for the cor.tainment wall were shown earlier (Fig. 3.9). The vertical displacement due to meridional strains is small (less than 3 inches) 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 defonnatio'n of containment would also be small. The nultiple penetration (X-71 and X-7.2) would not fail even for the most unfavorable forces which these pipes could sustain. For penetrations X-23 and 7
.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-r 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 median f ailure pressure for X-23 penetration, at which the leak areas shown n this table, is higher than the hoop failure pressure (216 psig) of the containment wall. These leak areas, therefore, are not expected to devel- , op. Penetration X-26 is expected to fail at a median pressure of 166 psig. The combined leak area for all safety injection penetrations is obtained by independently adding individual median leak area of 0.5 in2 , Penetrations X-71 and X-72 are not likely to contribute to the overall leak area, as stated earlier. The main feedwater lines (penetrations X-5 to X-8) are 18-in. diameter, Schedule 100 pipes. The failure mode of most concern is failure of the flued head due to axial loads in the pipe at the penetration. At a median pressure off 180 psig,2each one of these penetrations is likely to result in a leak area of 50 in each. Since all four2 of these can fail independent of each other, the total leak area is 200 in . Although the failure of a single such penetration can be considered as Type B failure, if all four main feedwater i penetrations were to fail simultaneously the resulting leakage will be of Type i C. .. i The fuel transfer tube is fixed to an elevated floor inside the contain-ment. As the pressure in the containment increases, the containment wall moves outwards and thereby exerts pressure on the bellows. The most pertinent l
i . . . ; 29 l 1 l Table 3.3 Leak Area Estimates for Mechanical Penetrations Median Median Line !
- Penetration Leak Area Failure Pressure Size .
in2 psig l X-20 to X-23 6.0 >216 12" j CCW Supply and Return [ X-24 to X-27 2.0 166 4" Safety Injection X-71 and X-72 Negligible i 1" Sample / Control X-5 to X-8 200 180 18" 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 k
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bellows from the viewpoint of containment leakage is the one inside the con-tainment (EP-2). Three potential failure modes, in their order of decreasing i probability of failure, considered are (a) failure due to overall buckling of the bellows, (b)-failure due to local buckling within the convolute, and (c) failure due to meridional 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 open to the contai nment atmosphere '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 contai ment. All eight valves are pneumatically operated butter-fly valves. At elevated temperatures, the seal material (usually ethylene propylene) on these valves may deteriorate and lose its sealing ' function. Any deposition of radioactive aerosols could further deteriorate the sealant materi al . Considering sealant degradation due to temperature alone, ethylene propylene seal life (Ref.10) is 5 hours, 40 mts, or 20 mts if exposed to 400, J 500 or 600 F, respectively. In the event of the failure of the sealant material, a narrow crack leak path may develop and containment atmosphere may being to leak into the space between the two isolation valves. Since the isolation 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-l lation butterfly valve also fails. The time duration elapsed before this I happens can be significantly long (of the order of hours). The SSPSA has es-timated it to be long compared to the containment failure by other causes. The SSPSA study, therefore, has disregarded this release path. The available leakage area due to sealant degradation has been estimated (Ref.10) by assuming an equivalent clearance of 1/16 inch between valve disc and body for ' low' and 1/8 inch for 'high' estimates. This gives a total leakage area of 17 in 2.as low value and 34 in as 2 high value. As noted ear-lier, the cuter butterfly valves must also experience high temperatures prior to a through release path. This leak area is of Category B. The SSPSA study has argued that such i 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-ting compound. The-SSPSA study has concluded that the failure of electrical penetrations 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 calculation , should be made to check this assessment. Such a calculation, similar to the problem of vent / purge line butterfly isolation valve failure, is beyond 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 sili. cone) . The structural failure, prior to containment failure, appears unlike-ly. The sealant material can degrade at high temperatures typical of a
severe accident. According to the 0-Ring Handbook (see Ref.10), silicone can survive for twenty hours when exposed to 500 F temperature. Furthermore, the personnel air lock is a double door system so even if the sealant around one door were to become ineffective, sbbstantial 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 B 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 the same pressure of 180 psig (195 psia).
Any depressurization due to a 50-in 2 hole is not likely to be fast enough to reduce the containment pressure substantially prior to the failure of the three remaining penetrations. Assuming that all four main feedwater lines fail at _180 psig, an equivalent leak area of 200 in2 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 likely to change. Estimated containment 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. Du ring normal 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 1 psid to 10 psid. In view of relatively strong primary containment, the role of the secondary containment is important primarily for Type B failures of the primary containment. 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. 3.3 Definition of Plant Damage States and Containnent 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 containment 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:
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- 1. Initiating Events "A" - Large Loss of Coolant Accident "S" -
Small loss of Coolant Accident "T" - Transient
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- 2. Timing of Core Melt and Conditions at Vessel Failure "E" -
No RWST Injection to RCS m' "L" With RWST Injection to RCS No Emergency Feedwater ( "F W" - No Emergency Feedwater i 3. Availability of Containment Systems "C" Long-Term Contaiment Spray Cooling "4" - Long-Tem Spray Recirculation, No Cooling "I" - Isolation Failure or 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. These damage states are categorized in a matrix of eight physical conditions in the containment (numerals (1) to (8)) and six combinations of containment safety function availability (letters A to F) for a total of 48 potential plant damage states. A ninth damage state type has been defined for accident sequences involving steam generator tube ruptures. Figure 3.11 indicates that only 39 plant damage states can be identified as credible sequences.
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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 f re-quency of each contaiment class is the sum of the frequencies of the plant states included therein.
- Annual plant state frequencies calculated by the applicant 5 for both in-
- ternal and external events were reviewed by the Lawrence Livemore National Laboratory 6 and were found acceptable. Table 3.6 presents the calculated contaiment class frequency estimates for internal events, fires, floods and truck crashes; rnderate 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 are initiated by earthquakes severe enough to impair evacuation. Fsr 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, floods and truck crashes. Seismic events with acceleration greater than 0.50g are judged to impair evac-uation, and must be treated separately in the consequence analysis. These contaiment response classes (or ant damage states) are the starting point for the containment event hree' nalysis and they define the link or interfaces with the plant analysis. 3.4 Contalment Event Tree and Accident Phenomenology An important step towards the development of the containment matrix in-volves the quantification of branch point probabilities in the containment event tree. These probabilities depend heavily on the analyses of degraded-and core melt phenomenology and the contaiment building response described in Appendix H of the SSPSA.5 l l
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/ $=>,/// $7afsom o45=GN FGafW E8 88s f Am Coott A Capaastify ges taPSA misa esoott Plant State Represents 2A AEC 4A TEC.SEC 2C/6C AE4 4C/8C TE4 10 AE 20/60 AL j 3D/70 SE,TE/TEFW 4D/80 SL,TL
, 2E/6E AECI 4E/8E TECl IF V j 2F/6F AEI 3F/7F SEI 4F/AF SLI 1 Figure 3.11 Definitions of the plant damage states used in SPSS.
. -3 5-Table 3.4 Frequencies of Occurrence of the Plant Dama9e States Frequency Frequency Plant Damage (events per Plant Damage (events per State reactoryear) State reactor year) 10 3.03(-7) 6A 3.41(-7) 1F 1.89(-6) 6C 3.57(-10) 1FA - 6.10(-11) 6D 2.49(-7) 1FP 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) 4D 2.79(-6) 9A 7.51(-10) 4E 2.24(-11) 9C 3.62(-13) 4F 2.25(-13) 90 4FP 9.09(-9) 1.18(-7)
TOTAL 2.30(-4) NOTE: Exponential notation is indicated in abbreviated form; i.e., 3.03(-7) = 3.03 x 10-7, i h ue-
Table 3.5 Containment Response Class Definitions Class Plant State Represents 1 ID AE 2 2A/6A, 4A/8A AEC, TEC, SEC 3 2C/6C 4C/8C AE4, TE4, SE4 4 3D/7D SE, TE, TEFW 5 2D/6D, 4D/8D AL, SL, TL 6 1F, 2F, 3F, 4F, 6F, .V 7F, 8F 7 2E/6E, 4E/8E AECI, TECI 4 8 1FP, 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) 4 9 V
- Table 3.6 Containment Class Mean Frequenciest Frequency (per reactor year)
Containment Internal, Fires,. Internal Response Class Floods and Truck Seismic <0.59 Seismic >0.59 Total Seismic and Crashes External 1 1.08E-7 - 1.95E-7 1.95E-7 3.03E-7 2 5.70E-5 1.54E-6 1.24E-6 2.78E-6 6.0E-5 3 1.80E-7 1.91E-8
- 1.91E-8 1.99E-7 4 8.60E-5 1.85E-6 2.27E-6 4.12E-6 9.0E-5 5 5.50E-5 1.10E-6 1.76E-6 2.86E-6 5.8E-5 6 1.80E-6 1.66E-7 3.93E-7 5.59E-7 2.4E-6 7 * *
- 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 * - - -
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- Indicates frequencies less than_10-8 yr 1 I
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The SSPSA containment event tree uses the twelve top events identified in Table 3.7 as major phenomena]ogical phases which could occur with respect to the formation and locat of core debris. These processes are grouped into four phases follow ng'and dent initiation (1) phenomena occurring while the core'is still' (2) phenomena occurring while the core is located below the lower gri ate but is still in the reactor vessel; (3 nomena occurring with the core debris located in the reactor cavity a onthe n-tainment floor; and (4) the phenomena involving long-tenn coolin co n-tainment and/or basemat penetration. 3.5 Containment Matrix (C-Matrix) J The twelve top events in the Seabrook containment event tree are summar-ized in Table 3.7. A negative response at any of the five nodes (4, 8,10, 11, and 12) in the containment event tree results in the failure of the con-tairrnent building by a variety of failure modes. Each of these failure modes results in a particular radiological release category. For those paths that do not have a negative response at any of the five nodes, the path will even-tually result in no failure of the containment. The containment 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 ends in a conditional probability (CP) of occurrence, and these cps should sum to unity. The quantification of an event tree is the process by which all the paths are combined to give the conditional probabilities of the various release categories. In SSPSA, fourteen release categories are used for the quantification as summarized in Table 3.8. Note that two of these release categories (namely, SS and S3) correspond to intact / isolated contai nment. Fission product release for this category would, therefore, be via nonnal leakage paths in the containment (and enclosure) building, which can befif----- S ferent-depending on availability of the enclosure building ventilatiogf fil- j teration system. ( / Table 3.9 sets forth a simplified containment matrix (C-matrix) for the Seabrook plant using the containment response class definitions discussed in Section 3.3, and the release category definitions given in Table 3.8. In arriving at the C-matrix of Table 3.9 all of the very low probability values were disregarded. This is shown7 to be insignificant to the risk estimate. The present assessment of containment response for Seabrook plant is not based upon independent confirmatory calculations of accident progression and containment response. Instead the knowledge gained from review of similiar risk studies for other 1 ,3," , pressurized water reactors with .large dry containments is used to guide this assessment. The mode and timing of containment failure cannot be calculated with a ' great degree of accuracy. Judgements must be made about the nature of the dominant phenomena and about the magnitude of several important parameters. Furthennore, the codes and methods used for these calculations are approximate and do not model -all of the detailed phenomena. Fortunately, risk measured in personal exposure is not sensitive to minor variations in failure mode and. timing. It is important, however, to properly characterize the major attri-butes of failure mechanisms; (1) whether the failure is early or late, (2) whether it is by overpressurization, bypass, or basemat melt-through and, (3) l whether or not radionuclide removal systems are effective.
=. -
Table 3.7 Accident Phase and Top Events for the Seabrook Containment Event Tree Accident Phase Top Event Initiator 1 Plant State Debris in Vessel 2 Debris Cooled in Place 3 No H2 Burn a 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 j(SmallLeak) i
~
7 r
. ) 1 Table 3.8 Release Categories Employed in the Seabrook Station Risk fiodel Release Category Release
- Group Category Definition S5 Containment intact / isolated with enclosure Containment air handling filtration working.
Intact / Isolated 3 S5 Same as S5 but with' enclosure air handling filtration not working. S2 Early containment leakage with late over-pressurization failure and containment building sprays working. W Same as S2, but with containment building spray not working. S2V Same as W, but with an additional vaporiza-tion component of the source term. S3 Late overpressurization failure of the con-Long-Term tainment with no ear'y 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 failure with containment building sprays working. Ni Same as S6, 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 explo-sion or hydrogen burn with containment building sprays working. E Same as S1, but with containment buildin~g
~
sprays 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.
t Table 3.9 Simplified Containment Matrix for Seabrook Release Category Containment Class S1 S2 S3 SS 56 il TI S2V TTi S4V S6V S6V-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 8 7 1.0 8 1.0 9 1.0 10 1.0 11 1.0 12 13 1.0 i I i i k 1
-42 .
The assessment of the contai nment 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 summarized as follows: Early Failure (S1,3T) which can result from a steam explosion or an early hy-drogen burn is believed to be unlikely. Although explosions in the reactor vessel lower plenum are highly probable, the resulting mechabal 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 have assigned a conditional probability of 10-4 to steam explosion induced containment failure. This probability leads 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 containment failure due to ex-ternal events (i.e., aircraft crashes) is assigned 1 in the SSPSA as shown in Table 3.6. This simply indicates that an aircraft crash into the containment is assumed to fail the containment structure with certainty. Early containment failure could also conceivably result from direct heat-ing 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 ll indicate that there exists a propensity for establishnent 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 contaiment by direct heat exchange and concomitant chemical reac-tions. Scoping calculations performed by the Containment Loads Working Group (CLWG) showed that a very severe challenge to the containment 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. Furthemore, 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 Lea'kage (S2, 37, 3N) without gross failure of containment building is expected to occur for nonisolated steam generator tube rupture event with containment sprays available (S2), for large break LOCA sequences
, with RWST injection in the abn nte of sprays (57) , and for dry cavity sequences with a vaporization release (S2V).
There seems to be a basic inconsistency in assigning plant damage states to this failure mode as defined in the C-natrix. Specifically, large brear LOCA sequences with RWST injection in the absence of containment sprays are expected to lead to an 3~3 failure mode with 100% probability (see 37 below); 1 l 1
while they are also assigned to '57 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 containment functions on steam generator
; tube rupture sequences is not at all obvious.
- Late Overpressurization Failure (S3, 57, T3V) can occur due to steam produc-
- tion in a wet cavity or noncondensable gas production as a result of core-con-crete interaction for a dry cavity situation. For sequences in which early and intemediate failure is not expected to occur, and for which containment sprays are inoperable, 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 overpassure 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 co ntai nment . l . 3.4 This is as a result of the very high failure pressure for the Seabrook con-tai ment. As a consequence of this high containment failure pressure (median pressure of- 211 for wet and 187 ~ psia for dry
- sequences) it is difficult to challenge the containment integrity by any conceivable event.
l l Hydrogen - deflagration ~ early in the accident sequence or later after l' 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 1 the containment integrity. The impact of changes in the contaiment failure distribution discussed in 3.2.5.4 is not significant for late failures. 1 Basemat Penetration Failure (S4, 3W) can only result in the absence of con- ] taiment heat removal system (sprays) for a dry cavity. A 26-inch high curb surrour.ds 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 r the basemat foundation. Therefore, the basemat penetration failure probabil-ities are conservatively assigned.
No Failure (55, '55) would result for all sequences with full spray operation. The radiological releases are thus limited to the design basis leakage with
; essentially negligible off-site consequences.
l Containment Isolation Failure (S6, FC is represented by an 8-inch diameter purge line. The accident sequences . where the containment is either not
~
i
*For dry sequences, only primary system water inventory is available in the contai nment. In this case, the containment atmosphere becomes superheated and,' at failure, the temperature can exceed 700*F.
i 1-_--.. , _ . - . - -.
isolated or bypassed (Event V) are assigned a conditional probability of unity to this release category. An interfacing systems LOCA (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-ment. Furthermore, the concurrent assumed loss of RHR and coolant make-up capability 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 injection lines 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 failing open while indicated closed" in the other motor-operated valve-in the nomal RHR hot-leg suction.
For the V-sequence, the core melts early with a low RCS pressure and a dry reactor cavity at vessel melt-through. The containment sump remains dry and recirculation is not possible. The core and containment 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 l ,3,4 for PWRs with large dry containments. Furthermore, the claimed unusually high strength of the Seabrook containment reduces the impact of sensitivity caused by uncer-tainties in the severe accident prog ression. However, should the claimed strength of the containment be reduced to levels comparable to some of the other large dry containments, the impact of uncertainties may become signifi-cantly more pronounced, as discussed in our review of the MPSS-3.7 3.6 Release Category Frequencies Based on the containment class frequencies in Table 3.6 and the contain-ment failure matrix of Table 3.9, the release frequencies were computed and are summarized in Table 3.10. 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 contaiment response classes and release categories, respec-tively. It is seen that containment classes 2, 4, and 5 dominate the core melt frequency while", the release categories SS (containment intact), U and S3V dominate the sou c tem frequency. 1
, Table 3.10 Frequency of Dominant Release Categories (yr-1)
Internal, Fires, Floods and Truck . Internal and Category Crashes Seismic.<0.59 Seismic >0.59 External S3 7.50E-7 3.45E-8 2.69E-7 1.05E-6 SS 5.64E-5 1.52E-6 1.23E-6 5.92E-5 lCE -
- 1.12E-7 2.40E-7 3.52E-7 lUi 5.50E-5 1.10E-6 1.76E-6 5.79E-5 S2V
- 5.29E-6 1.25E-5 1.78E-5 S3V 7.66E-5 1.65E-6 2.14E-6 8.04E-5 S4V 9.50E-6 2.04E-7 3.27E-7 1.0E-5 llgr 1.80E-6 1.66E-7 3.93E-7 2.36E-6 '
t 1
; Table 3.11 Contribution of Containment Response Classes to the Total Core Melt Frequency -
1 . Internal, Fires, Internal l Containment Floods and Truck- and j Class Crashes Seismic <0.5g Seismic >0.5g Total Seismic External 1 - - - -
<0.01 2 0.25 <0.01 <0.01 0.01 0.26 3 - - - -
<0.01 4 0.37 0.01 0.01 0.02 0.39 5 0.24 -
0.01 0.01 0.25 6 0.01 - - - 0.01 7- * * * *
- 8 0.025 0.055 0.08 0.08 9-13 1
2 ) i
- 1 .
Table 3.12 Release Category Frequency as a Fraction of Core Melt Frequency Release Internal Fires, Internal Category Floods and Truck and Crashes Seismic <0.5g Seismic >0.5g Total Seismic External l S3 <0.01 <0.01 <0.01 <0.01 <0.01 J S5 0.25 <0.01 <0.01 0.01 0.26 l0f *
<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 S3V 0.33 0.01 0.01 0.02 0.3S i S4V 0.04 <0.01 <0.01 <0.01 0.04 fl S6V 0.01 <0.01 <0.01 <0.01 0.01
,- 4. ACCIDENT SOURCE TERMS In this chapter the approach utilized in the SSPSA to detemine the 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 Assessment of Severe Accident Source Tems As in the Reactor Safety Study (RSS)l3 the CORRAL-II code was used in the SSPSA for detemining fission product leakage to the environment. This code takes input from the themal-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 failure, and core / concrete interaction. The methMs 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 determined ranging from contain-ment bypass sequence to the no-fail sequence as shown in Table 3.8. These release categories were evaluated by considering the containment f ailure mode, the avai. lability 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 contaiment, resulting from a steam explosion, early pressure spikes, or early hydrogen burns. Failure mode S2 represents a loss of containment functfori early in the accident sequence. This loss of function takes the form of an increase in the leak rate to 40% per day where it stays until the containment fails due to overpressurization. Failure mode S3 repre-sents a late overpressurization failure of the containment driven by decay heat or late hydrogen burns. Failure mode S4 represents a basemat melt-through, S5 represents no containment failure and the leak rate is limited to the contaiment design basis leak rate. Finally, failure mode S6 represents sequences where the containment is failed or bypassed as part of the initiat-ing event. The second parameter considered in defining the source tem is the avail-ability of sprays. This is determined by the plant damage states. Those
Table 4.1 Seabrook Point-Estimate Release Categories Time of Release Fractions by Group Initiation of Release Energy Release Seabrook Release Accident Release Duration Warning Time 6Release He19h t Ru La Category Sequence (hours) (hours) (hours)* 10 cal /sec (meters) Xe I-2** Cs Te Sa
< 10.0 10. .94 .023 023 .24 0033 41 9.8-5 51 AfC I.9 0.5 0.35 8.8-7 8.8-8 1.9 < 10.0 10. .89 2.1-5 2.1-5 4.4-6 2.9-6 52 AEC 2.6 1.0 1.3-8 3.8-9 3.8-10 66.1 0.5 62.5 210.0 10. .90 1.0-7 1.0-7 1.9-8 53 TE4
< 10.0 10 .0091 3.5-8 3.5-8 6.1-9 4.0-9 1.2-9 1.2-10 55 AEC 1.9 24 0.35 00044 .00013 1.3-5 4.0 < 10.0 10 90 .0036 .0036 00067 56 TEC1 4.5 4.0 210.0 10. 94 75 75 39 .093 46 .0028 3T AL 1.4 0.5 0.3
.15 092 .092 .017 011 0034 00034 W-1 7.3 9.1 6.2 < 10.0 10 00034
< 10.0 10. .24 .093 093 017 .012 .0034 U-2 20.3 IT. 19.2 .00046
< 10.0 10. .51 .12 .12 .023 .015 .0046 U-3 29.3 1.2 28.2 011 .0011
- 10. 90 .31 .31 .057 .038 H Total AL 7.3 27.3 6.2 --
i 26.4 210.0 10. 90 .122 .122 .022 .015 .0044 .00044 L 57 AL 27.2 0.5 6.3-8 1.9-8 1.9-9 ? 0.6 < 10.0 10. .014 5.2-7 5.2-7 9.5-8 55 TEC 4.3 24 l 2
< 10.0 10. .05 .037 .037 0069 0045 0014 .00014 WW-1 2.2 3.5 1.9 0079 0062 .0010
< 10.0 10 .10 .072 .072 0080 UV-2 6.2 7.2 5.9 022 .018 .0030 f
78.0 34.9 < 10.0 10 .85 .20 .20 .30 4 UV-3 35.2 .31 32 .034 .025 .0042 2.2 88.7 1.9 -- 10. 1.0 31 UV Total AE 210.0 10. 1.0 024 .024 .030 0026 0023 .00039 5TV TE 81.5 0.5 76.2 0062 0054 00091 2 210.0 10. 1.0 058 .058 072 5TV AE 50.0 0.5 49.6
.15 .11 .11 .020 .014 0041 00041 2.2 1.0 1.7 .35 10.
! 3TV-1 10. .31 .14 .14 026 .017 .0052 .00051 HV-2 4.2 3.0 3.7 .33 .0044 24 10 .51 .18 '.18 .36 .017 .024 HV-3 11.2 10.0 10.7 033 0053
.26 10. 97 43 43 40 .048 HV Total SE1 2.2 14.0 1.7 ft0TE: Esponential notation is shown in abbreviated form; i.e. 2.1-5 = 2.1 x 10 5 ,
*8ased on time of gap release except for $6 and Tsv where it is based on 0.5 hours af ter accident initiation.
" Elemental iodine - not used, all todine is treated as Csi.
1 o i i n- - - - l . .
release categories with operating sprayy systems are designated Si to S6, while those with spray systems not operating .are designated 3T to T6. , i The thi rd and final parameter considered in di f ferentiating 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 "V" in their designation as shown in Table 3.8. From the point of view of risk it was found 5 that 5F, ST, 537, and T6V 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 37f have late overpressurization failure modes, with no spray systems operating and differ only in the omission or inclusion of a vaporization release, respectively. The containment at Seabrook is cal-culated 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-ing in a puff release of approximately 0.5 hr release duration. From Table 4.1 it is seen that the 5T and 3Ti 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 case is due to the superior strength of the containment structure. Table 4.2 compares the 53, 5Tf release parameters with similar parameters for the other three reactors mentioned above. Note that a fair comparison should set (0!+1) equal to (Cs-Rb), since iodine was treated as Cs!. It is seen that I, Cs, and Ba groups for ST are apprcximately 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 sT7 release indicates a uniform 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 can also impact the radionuclide transport mechanisms inside the contain-ment due to changes in the containment themal hydraulic conditions. Release category 37f is associated with early containment f ailure 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 release up to the time when vaporization starts (melt + gap). The secon pus neludes the period of vaporization release and the third puff is equi to an overpressurization failure at the time of catastrophic containment failure. In this model the duration of the melt
Table 4.2 Late Overpressurization Failure Comparison 4 Millstone-37 Zion / Indiani " Indian 3 Seabrooks Point Study Point 37 T3V M-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) i 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) l Ba-Sr 1.5(-2) 2.6(-3) 3.0(-2) 3.7(-2) 2.5(-2) Ru 4.4(-3) 2.3(-3) 2.0(-2) 2.9(-2) 2.9(-2) 1 La 4.4(-4) 3.9(-4) l 4.0(-3) 4.9(-3) 1.0(-2) T (release) 27.2 81.5 20 (hrs) l T (duration) 0.50 0.50 0.50 0.50
! (hrs) i Energy 300E7 300E7 540E6 150E6 (Btu /hr)
I i f 1
release is seen to be 3.5 hours vaporization release 7.2 hours and the re-maining release 78.0 hours. It is not clear that the melt release in this case is 5.3 hours, however, it does not seem to be unreasonable. A 7.2 hour ' duration for .the vaporization release is not consistent with the RSS,13 which only allows 2 hours for this phase. Finally, it is not clear how the 78 hours for the last phase was determined. The release duration for a single puff, which is the sum of the above three phases leads to a release time of 88.7 hours which seems extraordinarily long. Our recommendation would be to reduce these times to be more consistent with RSS methods (see Table 4.7). The total release 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 5FT release is approximately half the M-4 release. 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 actual containment building. This feature is unique to the Seabrook contain-ment structure. Release category lI67 has binned into it an isolation failure correspond-i ng to an 8" diameter breach in contai nment and the interfacing LOCA (V-sequence). This sequence is also represented by a multi-puff release. In this case as in the previous case, the total release time is long compared to acceptable limits of the RSS 13 consequence model. Our recommendation would be to reduce these times to more reasonable values (see Table 4.7). The release fraction can be compared (Table 4.3) to the M-4 release from Millstone-3, PWR-2 for the RSS and the V-sequence from the RSSMAP study for Su 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 Uncertainty Analysis In this section we will briefly describe the uncertainty analysis carried out for the four dominant accident sequences and, where possible compare the fission product leakage to the environment to more mechanistic determina-tions. There are two contributors to the uncertainty in release characteriza-tion. First, uncertainty in time parameters which are influenced by:
- 1) Prediction of key event times, and
- 2) The mix of accident sequences binned into a release category.
Second, uncertainties in release fractions, which are influenced by:
- 1) Analysis methods and data, and
- 2) Uncertainties in timing of key events.
i i u
. Table 4.3 Comparison of Releases for Failure to Isolate Containment and the By-Pass Sequence Seabrook5 Millstone-37 RSS13* RSSMAPI s S2V 3RDI M-4 PWR-2 V-Sequence Xe 1.0 9.7(-1) 9.0(-1). 1.0 1.0 OI+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) 4 Te-Sb 3.2(-1) 4.0(-1) 5.0(-1) 3.0(-1) 4.4(-1) Ba-Sr 3.4(-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 (Btu /hr) 140E6 4E6 70E6 20E6 0.5E6
- *The same as M1A release category in Millstone-37 L _ _ _ _ _ . _ _. _ - -- - - -- - - - ~ -
The above principles were used to determine source term multipliers which would give a range of fission product leakage to the environment. A probabil-ity is associated with each source term, and for later overpressurization f ailure modes (53, 337, and 577) the following discrete probability distribu-tion is used, i.e.,
. Subcategory Probability 3T-a .02 3T-b .08 37-c .30 3T-d .60 This indicates, for example that there is an 8%' confidence level that 37-b correctly defines the source term for the 30i release category.
The results of this analysis for the overpressurization failure modes is: Particulate Release Factor (multiplier) Probability IT 3Tv 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 reicase, i.e., "d", the reduction factors of the source term 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 cont'ainment failures and are both binned into 53 and S3V sequences. 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 3]I and 51E 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 37 and STI releases. The release fraction determined for the TMLB' sequence is higher than all the 53 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 S3V. No comparative sequence for S2V was analyzed in BMI-2104. In the case of the 337 release category a different probability distribu-tion was used. This change reflects the break location, which initiates the i
\ , Table 4.4 Comparison of AB-c and TMLB'-c (BMI-2104) to S3V and U Release Fractions Release Probability Release Category Time (hrs) Xe Cs I. Te Ba Ru La S3V-a .02 28 1.0 _1.5(-2) 1.5(-2) 1.9(-2) 1.6(-3) 1.5(-3) 2.5(-4) 3Ti-b .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) , U-a .02 22 1.0 2.6(-2) 2.6(-2) 4.9(-3) 3.3(-3) 9.7(-4) 9.7(-5) E-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
l V-sequence. This break could be either in the hot-leg (b release subcategory) or the cold-leg (c release subcategory). This sequence is modeled as multi-puff release and each puff is treated separately. In this comparison only the sum 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 the various 55V releases and two V-sequence releases com-puted for Surry and published in BMI-2104. 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 for the " dry" V-sequence, are bracketed by the 33V 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-puff con-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 of the severe accidents based upon the RSS methodology. It must also be noted that the suggested source term for the Steam Gener-ator Tube Rupture (SGTR) sequence is assumed to be one-tenth of the source term for the event V (f67). 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 economic 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.
Table 4.5 Comparison of S6V (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) S6V-c .45 .97 1.295(-1) 1.295(-1) 3.2(-2) 1.593(-2) 5.2(-3) 5.3(-4) e C' , S6V-d .08 .97 5.2(-2) 5.2(-2) 1.3(-2) 6.6(-3) 2.0(-3) 2.2(-4) V (dry) - 1.0 5.52(-1) 1.99(-1) 1.2(-1) * *
- 1.0 * *
- V -
1.04(-1) 3.84(-2) 2.5(-2) .: (submerged)
- Individually not reported.
- , . l 1
. l Table 4.6 BNL-Suggested Source Terms Release Category Xe 01 I-2* Cs Te Ba Ru La S1. 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 S3 0.90 7E-3 1.E-7 1.E-7 1.9E-8 1.3E-8 3.8E-9 3.8E-10 S5 0.0091 - 3.5E-8 3.5E-8 6.1E-9 4.0E-9 1.2E-9 1.2E-10 S6 0.90 - 3.6E-3 3.6E-3 6.7E-4 4.4E-4 1.3E-4 1.3E-5 10I 0.94 - 0.75 0.75 0.39 0.093 0.46 2.8E-3 33i 0.90 - 0.31 0.31 0.057 0.038 0.011 1.1E-3 S2V 1.0 - 0.31 0.31 0.32 0.034 0.025 4.2E-3 1
! 30i 0.90 -
0.12 0.12 0.022 0.015 4.4E-3 4.4E-4 1 S3V 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 SS 0.014 7E-4 5.2E-7 5.2E-7 9.5E-8 6.3E-8 1.9E-8 1.9E-9 i 10fi 0.97 - 0.43 0.43 0.40 0.048 0.033 5.3E-3 33V-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 CsI.
**S6V-d release is 1/10th of the 3EV values.
i
! 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 i S2 2.6 1.0 0.5E6 10 1.05 S3 66.0 0.5 250E6 10 63 e 4 SS 1.9 10 n/a 10 0.35 l S6 4.5 4 0.5E6 10 0.50 33I 1.4 0.5 520E6 10 0.30 lUf 27 10 10E6 10 26 ! 155I 35 10 25E6 10 35 l 3CI 27 0.5 250E6 10 l 26 I 3Ti 81 0.5 450E6 10 76 { 3TV 50 0.5 250E6 0 49 1 s
- 55 4.3 24 10E6 10 0.30 ,
S6V 2.5 1.0 0.5E6 10 1.0 ! 337-d 2.5 1.0 0.5E6 10 1.0 :
- Warning time is defined as the time after core melt starts to the time of radiological release, i
i l i i 1 I I I i 1
, - + - - - , rrw-r---y---y v-w- ---wrw---wy --v--, -
.w-,y-==-r- yi-w =*-- --y v-rr-, r y-e =w .rvw--- , - - w --e ,i w -rm u-s%------
. c,.
1
- 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 performance, and radiological source tenn estimates for severe core melt accidents.
The containment response to severe accidents is judged to be an important f actor in mitigating the severe accident risk. There is negligible probabil-ity of prompt contaiment failure or failure to isolate. Failure during the first few hours after core melt is also unlikely. Most core melt accidents would be effectively mitigated by containment spray operation. Our assessment of the containment failure characteristics 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 used in 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 tenn study carried out by the Accident Source Tem Program Office (ASTP0) and reported in BMI-2104 it was found that the SSPSA releases were either higher than or for the most part similar to the recent release fractions. It was also found that the energy of release was somewhat higher in the SSPSA than for other existing studies. 1 7 O i . . . - 1 I
r
. m
- 6. REFERENCES
- 1. " Zion Probabilistic Safety Study," Commonwealth Edison Company (September 1981).
- 2. " Limerick Probabilistic Safety Study," Philadelphia Electric Co.
(September 1982).
- 3. " Indian Point Probabilistic Safety Study," Power Authority of the State of New York and Consolidated Edison Company (March 1982).
- 4. " Mill sto ne Unit 3 Drobabilistic Safety Study," Northeast Utilities (August 1983).
- 5. B. J. Ga rrick, et al . , "Seabrook Station Probabilistic Safety Assessment," Pickard, Lowe and Garrick, Inc., PLG-0300 (December 1983).
- 6. A. A. Garcia, et al., "A Review of the Seabrook Station Probabilistic Safety Assessment," Lawrence- Livermore National Laboratory Report (Dec.
12,1984). I 7. M. Khatib-Rahbar, et al., " Review and Evaluation of the Millstone Unit 3 J Probabilistic Safety Study: Contai nment Failure Modes, Radiological Source Terms and Off-Site Consequences," NUREG/CR-4143 ( report to be published). i 8. R. O. Wooten and H. Avet , " MARCH: Meltdown Accident Response Character-istics - Code Description and User's Manual ," BMI-2064, NUREG/CR-1711 (1980). 4
- 9. J. F. Muis, et al., "CORCON-Mod 1: An Improved Model for Molten Core / Concrete Interactions," SAND 80-2415 (1981).
4
- 10. B. E. Miller, A. K. Agrawal, and R. E. Hall, "An Estimation of Pre-Exist-ing Containment . Leakage Areas and Purge and Vent Valve Leakage Areas Re-sulting from Severe Accident Conditions," A-3741,11/15/84 (Draft report dated August 1984) transmitted via letter to V. Noonon, June 29, 1984
- 11. W. Lyon (organizer), "RCS Pressure Boundary Heating' During Severe Acci-dents," USNRC Meeting, Bethesda, Marylar.d (May 14,1984).
- 12. " Estimates of Early Containment Loads From Core Melt Accidents," Con-tainment Loads Working Group, NUREG-1079 (Draft 1985).
- 13. " Reactor Safety Study," U.S. Nuclear Regulatory Commission, WASH-1400, NUREG-75/014 (October 1975).
- 14. " Preliminary Assessment of Core Melt Accidents at the Zion and Indian Point Nuclear Power Plants and Strategies for Mitigating Their Effects,"
NUREG-0850, Vol.1 (November 1981).
- 15. G. S. Kolb, et al., " Reactor Safety Study Methodology Application Program: Oconee #3 PWR P1 ant," NUREG/CR-1659/2 of 4. j i
1 l
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