ML20140E047
| ML20140E047 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 02/28/1986 |
| From: | Ami Agrawal, Khatibrahbar, Ludewig H BROOKHAVEN NATIONAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20140E040 | List: |
| References | |
| RTR-NUREG-CR-4540 BNL-NUREG-51961, NUDOCS 8603270151 | |
| Download: ML20140E047 (68) | |
Text
i , ENCLOSURE
- s.
N'UREG/CR-4540 BNL/NUREG-51961 A REVIEW OF THE SEABROOK STATION PROBABILISTIC SAFETY ASSESSMENT: CONTAINMENT FAILURE MODES AND RADIOLOGICAL SOURCE TERMS M. Khatib-Rahbar, A. K. Agrawal, H. L' udewig and W. T. Pratt February 1986 Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 I
D D A
I.
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 I mitigating the consequences, (2) failure during the first few hours after l core melt is also unlikely and the timing of overpressure failure is very long '
compared to WASH-1400, (3) the point-estimate radiological releases are compa-rable in magnitude 'to those used in WASH-1400, and (4) the energy of release is somewhat higher than for the previously reviewed studies.
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CONTENTS Page ABSTRACT................................................................ iii ACKNOWLEDGMENT.......................................................... iv 1 LIST OF TABLES.......................................................... vi .
LIST OF FIGURES......................................................... vii
- 1. INTRODUCTION....................................................... 1 1.1 Background.................................................... 1 1.2 Objective and Scope........................................... 1 1.3 Organization of the Report.................................... 1
- 2. PLANT DESIGN AND FEATURES IMPORTANT TO SEVERE ACCIDENT ANALYSIS.... 2 2.1 Ar s es sment of Pl ant De si gn . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.2 Compa ri son wi th Oth er Pl ants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
- 3. ASSESSMENT OF CONTAINMENT PERFORMANCE.............................. 7 3.1 Containment Analysis Methods.................................. 7 3.2 Co n t a i nm e n t F a i l u r e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.2.1 Background............................................. 8 3.2.2 De s i g n De s c r i pt i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.2.3 Leakage Rate Calculation............................... 21 3.2.4 Containment Failure Model.............................. 22 3.2.4.1 Leak-Before-Failure........................... 22 3.2.4.2 Cl assi fi cati on of Fail ure . . . . . . . . . . . . . . . . . . . . . 23 3.2.5 Containment Pressure Capacity.......................... 24 3.2.5.1 Concrete Containment.......................... 24 3.2.5.2 Liner......................................... 27 3.2.5.3 Pe n et rat i on s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 3.2.5.4 Containment Failure Probability............... 31 3.2.5.5 Containment Enclosure......................... 31 3.3 Definition of Plant Damage States and Containment Response Classes.............................................. 31 3.4 Containment Event Tree and Accident Phenomenology............. 33 3.5 Contai nment Mat ri x ( C-Mat ri x ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 3.6 Rel ease Category Frequenci es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44
- 4. ACCIDENT SOURCE TERMS.............................................. 48 4.1 As sessment of Seve re Accident Sou rce Te rms . . . . . . . . . . . . . . . . . . . . 48 4.2 Source Term Uncertainty Analysis.............................. 52 4.3 R e c omme n d ed So u rc e Te rm s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56
- 5.
SUMMARY
AND' CONCLUSIONS............................................ 60
- 6. REFERENCES......................................................... 62 l
-vi-o .
LIST OF TABLES Table Title Page 2.1 Comparison of Sel ected Design Characteri stics . . . . . . . . . . . . . . . . . . . . 5 3.1 Containment Operating and Design Parameters......................
10 3.2 ~ Containment Liner Penetrations................................... 18 3.3 Leak Area Estimates fo Mechanical Penetrations. .. .... ... ... . . . . . . 29 3.4 Frequencies of Occurrence of the Plant Damage States............. 35 3.5 Containment Response Cl ass De fi ni ti ons . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 3.6 Containment Class Mean Frequencies............................... 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 Co ntai nment Mat ri x for Seab rook. . . . . . . . . . . . . . . . . . . . . . . 41 3,10 Frequency of Domi nant Rel ease Categories (yr 1) . . . . . . . . . . . . . . . . . . 45 3.11 Contribution of Containment Response Classes to the Total Core Melt Frequency.............................................. 46 3.12 Release Category Frequency as a Fraction of Core Melt Frequency........................................................ 47 4.1 Seabrook Poi nt-Estimate Rel ease Categori es . . . . . . . . . . . . . . . . . . . . . . . 49 4.2 Late Overpressurization Failure Comparison....................... 51 4.3 Comparison of Releases for Failure to Isolate Containment a nd t h e By- Pa s s Se q u e n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 4.4 Comparison of AB-c and TMLB'-c (BMI-2104) to 1"DT and 57. . . . . . . . .. 55-4.5 Comparison of S6V (sum) to V-sequence (Surry).................... 57 4.6 BNL-Su gge st ed So u rce Te rms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 4.7 BNL-Suggested Rel ease Characteri stics for Seabrook . . . . . . . . . . . . . . . 59 5.1 Comparison of SSPSA and WASH-1400 Containment Failure Frequencies...................................................... 61
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I LIST OF FIGURES Figure Titl e Page 3.1 A schematic representation of source term calculation............ 9 3.2 Equi pment hatch wi th personnel ai rl ock . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.3 Pe rs o n n el a i rl o c k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.4 Typical high energy piping penetration........................... 15 3.5 Typical moderate energy pi ping penetration. . . . .. . . .. . .. . . . . . .. . . 16 3.6 Typical electrical. penetration................................... 19 3.7 Typi cal ventil ati on penet rati on. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.8 A pictorial representation of leakage categories................. 25 3.9 Estimated radial di spl acement of containment wall .. .. .. .. . .. . .. .. 26 3.10 Estimated containment failure fractions.......................... 32 3.11 Defi nitions of the pl ant damage states used in SPSS. ... . .. . . .. . . . . 34 f
, 1. INTRODUCTION
1.1 Background
Probabilistic Risk Assessment (PRA) studies have been undertaken by a number of utilities (as exemplified by Refs.1-4) and submitted to the Nuclear Regulatory Commission (NRC) in support of various regulatory actions. Brook-haven National Laboratory (BNL) under contract to the NRC, has been involved in reviewing the accident sequence evaluations, core melt phenomenology, con-tainment response and site consequence aspects of the PRAs.
This report presents a review and evaluation of the containment failure modes and the radiological release characteristics of the Seabrook Station Probabilistic Safety Assessment (SSPSA), which was completed by Pickard, Lowe and Garrick, Inc. (PLG) for- the Public Service Company of New Hampshire and Yankee Atomic Electric Company in. December 1983.5 1.2 Objective and Scope The objective of this report is to provide a perspective on severe acci-dent propagation, containment response and failure modes together with radiol-ogical source term characteristics for the Seabrook Station. The determina-tion of accident initiation and propagation into core damage and meltdown.
sequences was reviewed 6 by the Lawrence Livermore National Laboratory (LLNL) under contract to the Reliability and Risk Assessment Branch of NRC.
In the present report, principal containment design features are discuss-ed and compared with those of Zion, Indian Point and Millstone-3 designs.
Those portions of the SSPSA related to severe accident phenomena, containment response and radiological source terms are described and evaluated. Numerical adjustments to the SSPSA 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 performance. Specifically, the definition of containment response classes and plant damage states, analytical methods, containment failure mod-el, containment event tree and accident phenomenology and the containment ma-trix are reviewed. Chapter 4 addresses the accident source terms together with justifications for adjustment where necessary. The results of this re-view are summarized in Chapter 5.
, , 2
, 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 identi fy commonalities for benchmark comparisons.
2.1 Assessment of Plant Design The Seabrook Station is comprised of two nuclear units each having an identical Nuclear Steam Supply System (NSSS) and turbine generator. The. units are arranged using a " slide.-along" concept which results in Unit 2 being ar-ranged 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 de-signed by United Engineers and Constructors.
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 containment, protects the liner over the foundation mat. .The contain-ment structure provides biological shielding for nomal and accident condi-tions. The aporoximate 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 ft. 6 1/8 in.
Foundation mat thickness 10 ft.
Containment penetrations are provided in the lower portion of the structure, and consist of a personnel lock and an equipment hatch / personnel lock, a fuel transfer tube, electrical, instrumentation, and ventilation penetrations.
Each containment enclosure (also known as secondary containment) 'sur-rounds a containment and is designed in a similar configuration as a vertical right cylindrical seismic Category I, reinforced concrete structure with dome and ring base. ~ The approximate dimensions of the structure are: inside diam-eter, 158 ft; vertical wall thickness, varies from 1 ft, 3 in, to 3 ft; and dome thickness, 1 ft, 3 in.
The containment enclosure and enclosure ventilation system is designed to collect leakage from the containment structure and to discharge the leakage to the filtration system of containment.* To accomplish this, the space between the containment enclosure and the containment structure, as well as the- pene-tration and safeguards pump areas, are maintained at a negative pressure fol-lowing a design basis accident by fans which take suction from the containment
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- Leakage via connections which pass into the steam and feedline penetrations are expected to bypass this system.
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enclosure and exhaust to atmosphere through charcoal filters. To ensure air.
~ tightness for the negative pressure, leakage through all joints and penetra-tions has bee.n minimized.
A containment spray system is utilized for post accident containment heat I removal ~. The containment spray system is designed to spray water containing i 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
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l j amount .of water is removed from the tank, the pump suction is transferred to j the containment sump, and cooling is continued by recirculating. sump water i through the spray heat exchangars and back through the spray headers.
i The spray-is actuated by a containment spray actuation signal which is l generated at a designated containment pressure. The system is completely re- -
dundant and is designed to withstand any single failure.
The containment isolation system establishes and/or maintains isolation '
of the 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-t
. curs or high. containment pressure is experienced. Low capacity thermal elec-
- tric hydrogen recombiners are provided.
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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-
- Jgin. The ECCS also provides continuous long-term-post-accident cooling of the
] core by recirculating borated water between the containment sump and the reac-i tor Core.
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- 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, i 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 3 purification and iodine removal, fresh air intakes, fire protection, emergency
] breathing apparatus, communications and meteorological equipment, lighting, j and housekeeping facilities.
. Engineered Safety Feature. (ESF) filter systems required to perform a safety-related function following a design basis accident are discussed below:
- a. The containment enclosure exhaust filter system for each unit col-l 'lects, filters and discharges any containment leakage. The system is j not normally in operation, but in the eventiof an accident, it is
. pl aced in operation and keeps the containment' enclosure and the 4
building volumes associated with the penetration tunnel and the ESF j equipment cubicles under negative pressure to ensure all leakage from i '
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the containment structure is collected and filtered before discharge to the plant vent. l l b. One of two redundant. charcoal filter exhaust trains is placed in op-eration 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.
j The emergency feedwater system supplies demineralized water from the con-
] densate water storage tank to the four steam generators upon loss of normal i feedwater flow to remove heat from the reactor coolant system. Operation of.
the system will continue until the reactor coolant system pressure is reduced l{ 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 I pump provides a diversity of power sources to assure delivery of condensate
-under emergency conditions.
j The two units of the facility 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 i 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 j loss of turbine generator power. In the event that the unit auxiliary trans-l 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, i diesel-engine generators. Either diesel-engine generator and its associated j safety feature bus is capable of providing adequate power for a safe shutdown i under accident conditions with a concurrent loss of off-site power. A con-t stant supply of power to vital instruments and controls of each unit is assur-j 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 the containment operating . pressure for Millstone-3 (subatmos-pheric design),- the use of fan coolers in Zion and Indian Point for post-acci-j dent containment cooling, lower reactor cavity configurations, and the chemi-cal composition of the concrete mix. The primary system designs are nearly i identical between the four units.
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! The Seabrook containment building basemat and the internal concrete j structures are composed of basaltic-based concrete. As concrete is heated, j 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- t i 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 l
)
Table 2.1 Comparison of Selected Design Characteristics ,
Zion Seabrook Design Parameters IndianPgint Millstony Unit 1 Unit 2 Unit 3 e Unit 1,2 Reactor Power IMW(t)I 3,250 3,030 3,411 3,650 Containment Bullding:
3 6 6 6 Free Volume (f t ) 2.73 x 10 2.61 x 10 2.3 x 10 2.7 x 106 Design Pressure (psla) 62 62 59.7 67.7 initial Pressure (psla) 15 14.7 12.7/9.1 15.2 initial Temperature (*F) 120 120 120/80 120 1
Primary Syste:
3 water Volume (ft ) 12,710 11,347 11,671 13,140
)
3 Steen volume (ft ) 720 720 7 2,012 Moss of UO2 in Core (lb) 216,600 216,600 222,739 222,739 l
Mass of Steel in Core (Ib) 21,000 20,407 7 19,000 Moss of Zr In Core (Ib) 44,500 44,600 45,296 45,234 Mass of Bottom Head (Ib) 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 Steam Generator System:
I inventory per Generator (Ibe) 77,000 82,000 113,600 112,000 Containment Buildino Coolers:
Sprays yes yes yes yes Fans (with safety function) yes yes no no Accumulator Tanks:
Total Moss of Water (lb) 200,000 173,000 348,000 213,000 Initisi Pressure (psla) 665 665 600 615
) Temperature (*F) 150 150 80 100-150 Ref uelino Water Storage Tank Total Mass of Water 6 6 6 (Ib) 2.89 x 10 2.89 x 10 10 2.89 x 10 Temperature (*F) 100 120 50 86 Reactor Cavity:
4 ConfIguratlon wet Wet Dry Dry / Wet Concrete Material Limestone Basaltic Besaltle Basaltle
' Minimum (Mexima Capacity = 3.9 x 10 6 lb) l 4
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amount of carbon dioxide.5 Thus, pressurization of the containment as a result of corium/cancrete interactions would be. expected to take a very long i time. In Table 2.1 the reactor cavity configuration is described as either
- wet or dry. Wet means that the reactor cavity configuration is such that for i ~ a wide range of accident sequences the cavity would be flooded. This means that if the reactor core melts down and falls into the cavity it would contact l
. significant quantities of water. This has important implications in terms of 1 l
ultimately cooling the co're debris and perhaps preventing core / concrete inter- l actions. It is also important in tenns of containment pressurization because !
boiling water by the ~ core debris is a much more efficient way of building up l pressure in containment than by core / concrete interactions. Dry in Table 2.1 i
implies that the reactor cavity configuration is such~that for a wide range of accident sequences the cavity would ~ not be flooded thus extensive core / con--
crete interactions would be expected.
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- 3. ASSESSMENT OF CONTAINMENT PERFORMANCE
,' In this chapter, the review of containment response to severe accider.ts is described. Analytical techniques used to analyze core meltdown phenomena
- and containment response are reviewed, the containment failure model is
- assessed, plant damage states and containment failure modes are evaluated.
, Parallels between this study and. other PRAs are made. Finally, the relevance .
and validity of the conclusions.is addressed.
! 3.1 Cop'cainment Analysis Methods
- A brief description of the computer codes used to perform the transient i degraded, core meltdown and containment response analyses is provided in this
! section.
- The MARCH 8 computer code is used to model the core and primary system l transient behavior and to obtain mass and energy releases from the primary.
system until reactor vessel failure. These mass and energy releases are then
! Jsed ' as input to the other computer codes for analysis of containment re-
- sponse.
i For. sequences in which the reactor coolant system remains at an elevated
! pressure until the vessel failure . (" time-phased dispersal"), the MODMESH S
} computer code is used. This code calculates the stean and hydrogen blowdown j 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 CORCON9 code is used to replace the INTER8 sub-3 routine of the MARCH code. CORCON models the core-concrete interaction after i the occurrence of dryout in the reactor cavity. The mass and energy releases i from the core-concrete interaction are transferred to the MODMESH code for j proper sequencing 5
and integration into the overall mass and energy input to
- C0C0 CLASS 9 code.
I C0C0 CLASS 9, a modified version of the Westinghouse COC0 computer code utilizes the mass and energy inputs to the containment 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 containment sprays and sump recirculation.
Fission product transport and consequence calculations are performed
! using the CORRAL-II and the PLG proprietary CRACIT S computer codes, respec-tively.
The analytical methods used to carry out the core and containment thermal hydraulics, and fission product transport calculations are identical to those used for MPSS-3.7 l
1
! 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 i
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-ure.
- . 3.2.2. Design Description The primary containment of the Seabrook plant is a seismic Category I re-il inforced concrete dry structure. It consists of an upright cylinder topped i with a hemispherical dome. The inside diameter of the cylinder is 140 feet j 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 j 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-4 sonnel airlock. The net free volume of the containment is approximately.2.7 x 106 ft 3.
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2 The inside of the containment is lined with a welded steel liner. The liner plate in the cylinder is 3/8" thick in all areas except penetration and i the junction of the _basemat and cylinder where it is 3/4" thick. This liner
} serves as a leak-tight membrane. Welds that are embedded in the concrete and i not readily accessible are covered by a leak chase system which permits leak i testing of these welds throughout the life of the plant. The dome liner is l 1/2" thick and flush with the outside face of the cylindrical liner. The op-l erating and the design parameters of containment are noted in Table 3.1.
- The containment building is surrounded by an enclosure. The containment enclosure is a reinforced concrete cylindrical structure with a hemispherical i
dome. The inside diameter of the cylinder is 158 feet. The vertical wall varies in thickness from 36 inches to 15 inches; the dome is 15 inches thick.
The inside of the dome is 5'6" 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.
The containment enclosure is designed to control any leakage from the 4
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.
l There are a number of containment penetrations which are steel components-that resist pressure. These penetrations are not backed by structural con-j crete and include the following: !
4
1 CONTAINMENT TIME OF FAILURE FAILURE' MODE WET OR DRY RELEASE SPRAY REACTOR CAVITY CATEGORY SYSTEM
~ SOURCE TERM O
Figure 3.1 A schematic representation of source term calculation.
Table 3.1 Containment Operating and Design Parameters Parameter Value -
Normal Operation .
Pressure , psig 0.5 Inside Temperature , F 120 Outside Temperature , F 90 Relative Humidity , % 45 Service Water Temperature , F 80 Refueling Water Temperature , F 86 Spray Water Temperature , F 88 Containment Enclosure Pressure , inches w.g. -0.25 Design Conditions Pressure , psig 52.0.
Temperature , F 296 Free Volume , ft 3 2.7x106 Leak Rate , % nass/ day 0.2 Containment Enclosure Pressure , psig -3.5
-11'-
- 1. Equipment hatch,
. 2. Personnel air 1ock,
- 3. Piping penetrations, j 4 Electrical penetrations,
- 5. Fuel transfer tube assembly,
- 6. Instrumentation penetrations, and l 7. Ventilation penetrations. ,
l These components penetrate the containment and containment enclosure shells to j provide access, anchor piping, or furnish some other operational requirement.
l 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 l} 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) l Barrel 3 1/2 i Spherical 1 3/8 Flange 5 3/8 Air lock mounting 1 1/2 sleeve
! The equ.ipment hatch cover is fitted with two seals' that enclose a space j which can be- pressurized to 52.0 psig. The flange of the cover plate is at-l tached to the hatch barrel with 32 swing bolts,1-3/8 inch in diameter. The l 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:
Parameter Dimension Inside Diameter of Barrel 9'6" l Barrel Thickness 1/2" Door Opening 6'8" x 3'6" Door Thickness 3/4" Bulkhead Thickness 1-1/8" I Each door is locked by a set of six latch pin assemblies, and is designed to j 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 i
and both swing into the containment. Each door is fitted with two seals that 1
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u Figure 3.2 Equipment hatch with personnel airlock. -
O e
4
! 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 '
l j ' pressure by use of special tools. The doors may. be operated mechanically. ;
l 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" i Barrel Thickness - 5/8" Cover Thickness ,
'5/8" 3 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 ring, breach-type mechanism. These doors are also mechanically interlocked so that only one door can be opened at a time. The capabilityEexists for bypassing this interlock and relieving the internal pressure by use of special tools.
The doors may be operated mechanically.
f Piping Penetrations 1
There are two types of piping penetrations: moderate energy and high energy. 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.
4 High energy piping penetrations (Fig. 3.4) consist of a section of pro-
! cess pipe with an integrally-forged fluid head, a containment penetration i 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 ration between the pipe and the support. The annu-i lar space between the process pipe and the sleeve is completely filled with fiberglas; thermal insulation. The pipe and the fluid head, are classified as i ASME -III Safety Class 2 (NC), whereas the sleeve is classified as part of the i 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-j pl ate . The pipe is classified as ASME III Safety. Class 2 (NC). The sleeve is classified as ASME III Div. 2 (CC). The end-plate is classified as Class MC.
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l l i Table' 3.2 gives a list of the containment piping penetrations. Included i' 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 terminal box, electrical modules which are clamped to the header plate, and a' carbon steel weld ring whichiis. welded to the header plate and to the sleeve. The metallic pressure resisting parts, the sleeve, I -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 i backed by concrete was designed as part of the concrete containment, ASME III (CC).
Double silicone and Hypalon 0-rings provide a seal with a . cavity for leakage monitoring between the header plate and the modules. The header plate is provided with a hole on the outside of the 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 The fuel transfer tube assembly consists of the fuel transfer tube, the penetration sleeve, the fixed saddle on the reactor' side, and the sliding sad-die in the fuel storage building. The fuel transfer tube centerline is at l 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 j an inside diameter of about 25".
Ventilation Penetrations
! There are two types of ventilation penetrations - the containment air
, purge penetrations (HVAC-1 and HYAC-2) and the containment on-line penetra-i tions (X-16 and X-18). The containment 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.
I' Together with the pipe, these valves form a part of the containment pressure boundary. The valves are 36" diameter butterfly valves with fail-safe-pneu-
- matic operators. The weld between the pipe and the containment 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 wel.ded to the sleeve ~on the' outside of the containment, and a 3/4" valve and-test connection is attached to it. The i .
i
~.s -, n . _ . . , . _ . .,.%o __,-,--_,,,,,,.7 ,__,,e 4__,,yc pw,p, , ,, - .,p,.
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" 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 i 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 <1" X X-60, X-61 From containment recirculation sump 16" X-62 Fuel transfer tube 20" X-63 to X-66 Steam generator blowdown 3" X-67 Service air 2" -
HVAC-1,2 Containment purge supply / exhaust lines 36" X-19,~X-41, X-42 X-44 to X-46, X-48 Spare ?
X-49, X-51, X-58 X-59, X-68 to X-70 l
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OUTBOARD CONTAINMENT WALL INBOARD STAINLESS STEEL LINER PLATE HEADERPLATE]
THERMAL INSULATING GASKET-Y 1-BOX MOUNTING RING CANISTER J_ -
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CARBON STEEL WELD RING
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- FIELD WELD SLEEVE /
JUNCTION (TERMINAL) BOX JUNCTION BOX Figure 3.6 Tyical electrical penetration.
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r Figure 3.7 Typical ventilation penetration.
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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 sleev'e 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 builds up 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 2 k-1 W= k(kT1) A8 (1) where W = discharge rate (kg/s),
A = leak area (m2),
P = absolute pressure (N/m2),
o = 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 o= (2) where R = gas constant, and T = the absolute temperature, Then Eq.(1) becomes k+1 W= k(k+1) 2 k-1 [P A V RT (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 k-1 f= k(k 1)- MA (5)
Note that the leakage rate, when expressed in terms of mass fraction, depends only on the leakage area.
For Seabrook-1, using V = 2.704x106 ft3 and T = 296 F Eq.(5) gives Leakage Rate = 0.721 Ain w/o per hour (6) where Ain is the leakage area in in2 . Al ternately, Leakage Rate = 17.3 Ain w/o per day. (7)
The essentially intact containment leakage of 0.2 w/o per day, thus, corre-sponds to an equivalent leakage area of 0.012 inz (or, an equivalent hole of 1/8-in diameter). A leakage area of 4 to 10 in2 would-correspond to the leak-age rate of 2.9 to 7.2 w/o per hour. In other words, it will take about 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to leak the entire content to the environment through a 10-in2 hole.
3.2.4 Containment Failure Model 3.2.4.1 Leak-Before-Failure During accident sequences involving core damage, the containment struc-ture will be exposed to pressures and temperatures 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 termed local failure. The containment leakage can occur at many locations and discontinuities such as mecharical 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 termed a threshold model. In the threshold model, tie 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- l 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
.l and thus a catastrophic or gross failure may be avoided. '
~23-I 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 develop a small . leak that is substantially larger than the leak ac-ceptable from an intact containment, but not large enough to arrest the pressure rise in the containment.- Category A failures thus cause an early increase in the rate .of leakage of radionuclides over the de--
sign basis leak rate but-pressurization of the containment' continues until either a category B or C containment failure occurs.
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-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 j 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 -
j able leakage to be quoted as w/o per day. The higher value noted here i
is based on Amendment 53, August 1984.*
t
! . Containment Failure Category B. Includes failure modes that develop a j large enough leak area so that the pressure in the containment no i
longer increases. The time during which a substantial fraction of the' radionuclide source term is released is longer than approximately 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Category B failures include self-regulating failure- modes l, where the leak area is initially small but increases with pressure so 1 that it _becomes sufficient to- terminate the pressure rise before a category C containment failure occurs.
The definition of " substantial" fraction is unclear.
. Containment Failure Category C. Includes those containment failure
- modes that develop a -large leak area. A large fraction of the total i
radionuclide source ' term is rele; sed over a period of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. All gross failure modes are included in category .C.
l
- Mathematically, these three failure categories can be expressed in ' terms of leakage areas as follows:
j Type A ADBA < AA i ANP ANP < AB i Ap Type B (8) ,
AC > Ap Type C where
- ADBA = leakage area corresponding to the technical specification limit for ' containment leakage, i *There appears. to be substantial update / changes in the Engineered Safety
{ Features flow diagram, including arrangements of motor operated valves an.d j bypass lines, ~which may substantially change the frequency of events. ~ BNL, however, is not Lreviewing this part of SSPSA.
1 1
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, - . , _ - - - - , - , _ , ..,,.___.--_.,._--__-.....,.,,,-.....--_.-A,.-....,_,,...._..J_.---.,-.---,---
ANP = leakage area not large enough to arrest pressurization, and Ap = leakage area sufficient to release 100 w/o in one hour.
The leakage area required to release a substantial fraction of the radio-nuclide source term in approximately an hour can be computed using Eq. (6).
Assuming one-hundred percent turnover as " substantial" fraction in one hour, Eq. (6)- gives the required leakage area to be equal to 138 in 2 or about 1 ft 2. If one were to use 50v, turnover as " substantial" in an hour, the corre-sponding leakage area would be 60 in 2 or about 0.5 ft 2. Forthepgrposeof defining an upper bound for type B, it seems justified to use 1 ft , hence, leakage area in excess of this value should be considered type C failure Al-though this estimate of the leak area is a factor of two too high from the value stated in SSPSA, it is not significant in the risk evaluation since failure categorization is somewhat arbitrary.
The leakage area required to arrest containment pressurization is in the range of 4 to 10 square inches. A leak area of about 6 square inches will re-sult in the release of about 100 w/o of activity in a dag (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 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The Category B leak area is, thus, in the range of 4 in2 to 1 ft . Figure 2
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 fot.ad 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, thus, assumed to fail at this pressure. At pressures.beyond this, very large irreversible deforma-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 deformations of the containment wall are shown in Figure 3.9.
Note that the radial strain at the expected failure pressure of 216 psi is 4.71, ( 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
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I Figure 3.8 A pictorial representation of leakage categories..
t i I I I 1
157 (yield) 180 200 Failure 216 20
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RADIAL DISPLACEMENT OF THE CONTAINMENT WALL AWAY FROM THE BASE (INCHES) . .
Figure 3.9 . Estimated radial displacement of containment wall.
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i 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 an estimation of leak areas was, therefore, not considered for this mode of
. failure.
i 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 i 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 l j psi is reached. Thus, the failure of the base slab and containment wall junc- 1
- tion is not limiting.
! In summary, the containment wall is expected to undergo significant de-i formation (=4.7% ar/r) prior to its failure at 216 psig. At this pressure, i Type C (i.e., gross) failure occurs.
! 3.2.5.2 Liner
-The elongation capacity of the steel liner is computed by neglecting the j friction forces between the liner and the concrete. The possibility that the 4 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 unaxial conditions, or an elongation of 4.7 percent under plane i
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.
1 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.
4 i As the containment pressure increases beyond its yield value (157 psi),
4 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:
4 Penetration X-23 12" schedule 40 carbon steel (also X-20 to X-22 by
- similarity)
) Penetration X-26 4" schedule 160, stainless steel j (also X-24, X-25, X-27) m - .. -,e.- -c..- ~,-,,,.,.w -
, , . - , .~w -,--..,.,,.n-,, , . , , , . - - - , ,- ..- n - - -,n--.-, ,-,,---n.,~-....v,. , , -
l 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 containment 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 deformation of containment would also be small.
The multiple penetration (X-71 and X-72) would not fail even for the most
- unfavorable forces which these pipes could sustain. For penetrations X-23 and X-26, the most likely location for failure is at the partial penetration fil-let welds which join the pipe to the end plate. When failure of this weld oc-curs, the pipe remains in the hole provided in the end plate. The gap between the pipe and the end plate is likely to remain small unless the pipe wall buckles. Exact gap size is hard to compute. The SSPSA appears to use a uni-form gap size of 0.04 in., and 0.10 in. as median and upper estimates, respec-tively. The corresponding leak areas for X-23 (as well as X-20 to X-22) and X-26 (as well as X-24, X-25, and X-27) penetrations are shown in Table 3.3.
The median failure pressure for X-23 penetration, is higher than the hoop failure pressure (216 psig) of the containment wall. These leak areas, there-fore, are not expected to develop.
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 hrid due to axial loads in the pipe at the penetration. At a median pressure of 180 psig, each one of these penetrations is likely to result in a leak area of 50 in2 each. A leakage area of 50 in2 , either from the failure of one or more feedwater line(s), will result in a substantial pressure reduction and l thereby further deterioration of leakage area may not materialize. Neverthe-less if all four lines were to fail simultaneously, total leakage area of 200 in2 may result. In this case it will be category C failure.
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 1
- . _ , _ _ . . - . _ . . , , _ _ _ . - - ,__e. _ , _ _ , . _ - . - _
Table 3.3 Leak Area Estimates for Mechanical Penetrations Median Median Line Penetration Leak Area Failure Pressure Size in2 psig X-20 to X-23 6.0 >216- 12" CCW Supply and Return X-24 to X-27 2.0 166 4" Safety Injection 4
X-71 and X-72 Negligible i 1" Sample / Control X-5 to X-8 50 to 200 180 18" l 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 i
4 I
bellows from the viewpoint of containment leakage is the one inside the con-tainment (EP-2). Three potential failure modes, in their order of decreasing probability of failure, considered are (a) failure due to overall buckling of the bellows, (b) failure due to local buckling within the convolute, and (c) f ailure due to meridional bending strains. The SSPSA has estimated median leak' area of about 3 inz 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 containment 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 containment. 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 material. . Considering sealant degradation due to temperature alone, ethylene propylene seal life (Ref.10) is 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, 40 mts, or 20 mts if exposed to 400, 500 or 600 F, respectively.
In the event of the failure of the sealant material, a narrow crack leak path may develop and containment atmosphere may begin 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 i
to the environment can occur only if the sealant of the outer containnent iso-lation butterfly valve also fails. The time duration elapsed before this
- happens can be significantly long (of the order of hours). The SSPSA has es-timated . it to be long compared to the containment failure by other causes.
The SSPSA study, therefore, has disregarded this release path.
The available leakage area due to sealant degradation has been estimated (Ref.10) by assuming an equivalent clearance of 1/16 inch between valve disc and body for ' low' and 1/8 inch for 'high' estimates. This gives a total leakage area of 17 in z as low value and 34 in2 as high value. As noted ear-lier, the outer butterfly valves must also experience high temperatures prior to a through release path. This leak area is of Category B. The SSPSA study i has argued that such a leak path is not likely to result prior to a gross containment failure (Category C).
Electrical penetrations can fail primarily due to overheating of the pot-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 thermal 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 t
i
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, substantial time delay would be required to make the second sealant also ineffective. It, thus, appears that the equip-ment hatch and personnel lock penetrations do not contribute significantly to Type B failure.
3.2.5.4 Containment Failure Probability The calculation of the probability of containment failure as a function of the pressure is quite involved. The method used and results reported in the SSPSA study seem reasonable except for the impact of all four main feed-t 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 in2 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-in2 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. Use of the modified failure fractions do not seem to cause any appreciable change in the risk estimates.
3.2.5.5 Containment Enclosure i
The containment enclosure building is designed to withstand 3.5 psi pres-sure difference between the enclosure and the environment. During 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. The SSPSA study, however, has not taken any credit for -
, the enclosure building.
3.3 Definition of Plant Damage States and Containment 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:
- 1. Initiating Events l "A" -
Large loss of Coolant Accident "S" -
Small loss of Coolant Accident "T" -
WET SEQUENCES 1.0 _ ,_ , , , ,
g - - - SSPSA i
- BNL 0.8 -
\
<c Benign
$ (Type B) \
j u failures N
< N.
E 0.6 - ~ -
E!
5 Q
w -
i Z l r 0.4 - - - -
z -
q Gross /
g (Type C) /
g Failures N i
0.2 -
/ -
./
i i l I i ! .
- 120 140 160 180 200 220 240 1
PRESSURE, PSIA Figure 3.10 Estimated containment failure fractions. .
l
~
G
- 2. Timing of Core Melt and Conditions at Vessel Failure
, "E" - No RWST Injection to RCS "L" - With RWST Injection to RCS
""- - No Emergency Feedwater "FW" - With Emergency Feedwater 1
- 3. Availability of Containment. Systems ,
i "C" - Long-Term Containment Spray Cooling "4" - Long-Term 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.
From the viewpoint of containment response, many of the plant damage states can be grouped into containment classes. The classes defined in Table 3.5 are differentiated primarily according to spray availability. The fre-quency of each containment class is the sum of the frequencies of the plant.
states included therein.
Annual plant state frequencies calculated by the applicant 5 for both in-ternal and external events were reviewed by the Lawrence Livermore National Laboratory6 and were found acceptable. Table 3.6 presents the calculated con-tainment class frequency estimates for internal events, fires, floods and truck crashes; moderate and severe seismic events.
In order to comprehensively assess the risk from seismic events, it is necessary to make separate consequence calculations for those accidents which are initiated by earthquakes severe. enough- to impair evacuation. For this purpose, the seismic frequency estimates are divided into two categories in Table 3.6. The seismic events with instrument peak ground acceleration below 0.5g 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 containment response classes (or plant damage states) are the starting point for the containment event tree analysis and they define the link or interfaces with the plant analysis.
3.4 Containment 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 containment building response described in Appendix H of the SSPSA.5
. .- . . , - . - - - - , ~ - - . . , . =
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Plant State Represents 2A AEC
. 4A TEC,SEC i 2C/6C AE4
- 4C/8C TE4 l
10 AE .
! 40/80 SL,TL 2E/6E AECI 4E/8E TECI
- 1F V 1
2F/6F AEI 3F/7F SEI 4F/8F SLI
{
Figure 3.11- Definitions of the plant damage states used in SPSS.
e
-n, r -,-,. e ,.m,,,.e,.,. -.--,-,---g-.v,-, -,-.~,---~,,,,,.<.-,e,.-n-,m.,~,.--n-. 7-.---,--w-,-e n.- -,-,w-.,. -,+--ees,,-~, .--,,,w--. , - - - , - - -
. Table 3.4 Frequencies of Occurrence of the Plant Damage States Frequency Frequency l
Plant Damage (events per Plant Damage (events per .
)
State reactor year) S_ tate reactor year) i 10 3.03(-7) 6A 3.41(-7) i 1F 1.89(-6) 6C 3.57(-10) 2.49(-7) 1FA 6.10(-11) 6D 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).
i 2FP 1.58(-10) 8A 4.50(-5) 30 1.94(-5) 8C 4.29(-8) 3F 5.00(-7) 80 5.51(-5) 3FP 6.21(-6) 8E 5.02(-11) 4A 1.28(-5) 8F 1.02(-10) 4C 1.65(-7) 8FP 1.95(-7) 40 2.79(-6) 9A 7.51(-10) 4E 2.24(-11) 9C 3.62(-13) 4F 2.25(-13) 90 9.09(-9) 4FP 1.18(-7)
TOTAL 2.30(-4)
NOTE: Exponential notation is indicated in abbreviated form;
- i.e., 3.03(-7) = 3.03 x 10-7
! l 1
i
l Table 3.5 Containment Response Class Definitions Class Plant State Represents 1 1D l ~AE
- ~
2 2A/6A, 4A/8A AEC, TEC, SEC 3 2C/6C, 4C/8C AE4, TE4, SE4 4 30/70 SE, TE TEFW 5 20/60, 40/80 AL, SL, TL 6 1F, 2F, 3F, 4F, 6F, V
- 7F, 8F 4
. 7. 2E/6E, 4E/8E AECI, TECI J
8 IFP,3FP/7FP Small leaks w/o RWST 9 2FP/6FP, 4FP/8FP Small leaks w/ RWST
! 10 1FA, 2FA/6FA Aircraft ' crashes I
11 9A V2 (SGTR) 4 12 9C V2 (SGTR) 13 90 V2 (SGTR) 4 5
i t
i
.~
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 i
1 1.08E-7 -
1.95E-7 1.95E-7 3.03E-7 2 5.70E-5 1.54E-6 1.24E-6 2'78E-6
. 6.0E-5 1.80E-7 1.91E-8
- 3 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 * * * * *
- 5.29E-6 1.25E-5 8 1.79E-5 1.79E-5 e 9
- 1.12E-7 2.40E-7 3.52E-7 3.52E-7 U' 10 * -
11 * - - -
- l 12 * - - -
- 13 * - - - *
- tReference [5] Tables 5.1-3 and 9.2-9.
~
- Indicates frequencies less than 10-8 yr l.
S 1
9
The SSPSA containment event tree uses the twelve top events identified in Table 3.7 as major phenomenological phases which could occur with respect to the formation and location of core debris. These processes are grouped into four phases following an accident initiation (1) phenomena occurring while the core is still in place; (2) phenomena occurring while the core is located be-low the lower grid plate but is still in the reactor vessel; (3) phenomena oc-curring with the core debris located in the reactor cavity and on the contain-ment floor; and (4) the phenomena involving long-term cooling of the contain-ment and/or basemat penetration.
3.5 Containment Matrix (C-Matrix)
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-tainment 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 li_nks 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 re-
. lease 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, S5 and SS) correspond to intact / isolated containment.
Fission product release for this category would, therefore, be via normal leakage paths in the containment (and enclosure) building, which can be dif-ferent depending on availability of the enclosure building ventilation / fil-tration 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 was 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 similar risk studies for otherb3." pressurized water reactors with large dry con-tainments 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.
Furthermore, the codes and methods used for these calculations are approximate and do not model all of the detailed phenomena. Fortunately, risk mtasures are not sensitive to minor variations in failure mode and timing. It is im-portant, however, to properly characterize the major attributes of failure
, mechanisms; (1) whether the failure is early or late, (2) whether it isby overpressurization, bypass, or basemat melt-through and, (3) whether or not radionuclide removal systems are effective.
Table 3.7 Accident Phase and Top Events for the
~Seabrook Containment Event Tree i
! Accident Phase Top Event .
Initiator 1 Plant State Debris in Vessel 2 Debris Cooled in Place t 3 No H2 Burn 4 Containment Intact Debris in Reactor Cavity 5 Debris Dispersed from Cavity
- 6 Debris Cooled 7 No H Burn 8 Containment Intact Long-Term Behavior 9 No Late Burn 4
10- Containment Shell Intact 11 Basemat Intact Failure Mode 12 Containment Failure
- l *Both event tree paths are containment failure; that is, success is small leakage and failure is gross leakage.
]
J.
4 I
4 l.
4 l 1
I I
j ,
4 I
Table 3.8 Release Categories Employed in the S'eabrook Station Risk Model i
- Release Category Release
- Group Category. Definition ;
S5 Containment intact / isolated with enclosure '
l Containment .
air handling filtration working.
i i
Intact / Isolated .
SS Same as S5 but with enclosure air handling i filtration not working.
}
} - 52 - Early containment leakage with late over-i pressurization failure and contair.. tent ,
l building sprays working. '
H Same as S2, but with containment building-l spray not working.
i
, W Same as W, but with an additional vaporiza-l tion component of the source term.
i S3 Late overpressurization failure of the con-i Long-Term tainment with no early leakage and contain-4 Containment ment building sprays working, i Failure i j TJ Same as S3, but with containment building i
- sprays not working.
TJi Same as TJ, but with an additional vaporiza-j tion component of the source term.
j S4 Basemat penetration failure, sprays operating
)j ' T4Y Containment basemat penetration failure with containment building sprays not working and additional vaporization component of the j
source term.
j - ;
, S6 Containment bypass or isolation failure with '
l containment building sprays working.
I
! T67 Same as 56, but with containment building j sprays not working and an additional vapori-1 Early zation component of the source term.
. Containment
} Failure / Bypass S1 Early containment failure due to steam explo-i sion or hydrogen burn with containment j building sprays working.
1 i
W Same as 51, but with containment building sprays not working, i "5 denotes applicability to Seabrook Station; number corresponds with contain l
ment failure mode; bar denotes nonfunctioning of containment building sprays; i
! and V denotes achievement of sustained elevated core debris temperatures and associated vaporization release. '
.l s
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~
Table 3.9 Simplified Containment Matrix for Seabrook Release Category Containment Class S1 S2' S3 SS S6 Tl' 37 T2V TW T4V TW - T67-d -
1 0.60 0.40 2 0.01 0.99 3 1.0
- 4 0.89 0.11 ,
i 5 1.0
- 6 1.0 3
7 1.0 8 1.0 9 1.0 i
j 10 1.0 J
1 11 1.0 12 1.0
, 13 '1.0 I
I
! 1 J
s k
2 l
I I
l
,A
.- __ _ .._ _m . . _ _ _ . __ . - _ _ ,
i .
The assessment of the containment response and failure mechanisms is
' based on the general understanding of that accident phenomenology and the con- -
i tainment design characteristics discussed earlier. The phenomena of interest may be sumarized as follows:
l-
{ Early Failure (S1, TT) 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 probable, the resulting_ mechanical energy- would be
- limited by the fraction of the core which could participate in a single explo- t i sion and by the efficiency of the - process. In recent PRA reviews,4,7 we j 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 4 value is not included in the simplified C-matrix
! The conditioral probability for an early containment failure due to ex-1 ternal events (i.e., aircraft crashes) is assigned 1 in the SSPSA as shown in i Table 3.9. This simply indicates that an aircraft crash into the containment l,
1s assumed to fail the containment structure with certainty.
! Early containment failure could also conceivably result from a rapid dis-4 persal of the core debris throughout containment in the form of aerosols which
) directly heat the atmosphere producing a rapid pressure / temperature pulse.
i The dispersal could only be caused by the high primary system pressures that 1
may exist at vessel failure for a number of transient sequences (recent calcu- '
lationsll indicate that there exists a potential for the establishment of a j 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 i
depressurization prior to bottom head melt through, thus eliminating, ' high pressure ejection sequences). The aerosols could rapidly pressurize contain-
- ment by direct heat exchange and cor. comitant chemical reactions. Scoping cal-i culations performed by the Containment Loads Working Group (CLWG) showed that a very severe challenge to the containment integrity could result provided 25 i percent of the core mass were converted to aerosols.12 However, no consensus l coula be reached among the CLWG analysts as to the credibility of this param-eter value, and this failure mode is still speculative. Furthermore, the con-j j figuration of the Seabrook lower cavity would tend to reduce the dispersal of
! core debris beyond the reactor cavity boundaries. .
j For the reasons outlined above (as well as the high containment failure i pressure for Seabrook), it is concluded that early overpressure failure has a j very low likelihood.
i
{ Early Containment Leakage (S2, 5T, T2V) without gross failure .of containment j building is expected by the applicant to occur for large break LOCA sequences l with RWST injection in the absence of sprays (T2), and for dry cavity
] sequences with a vaporization release (T2V).
- The plant damage states for steam generator tube rupture sequences are j assigned to the S2 release category without any attempt to compute a SGTR l specific release category.
4 i
Late Overpressurization Failure (S3, 57, TN) can occur due to steam produc-tion in a wet cavity or noncondensable gas production as a result of core-l'
concrete interaction for a dry cavity situation. For sequences in which early and intermediate 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 for large break LOCAs and 0.89 for small breaks and transtants. This results from the relative com- '
petition' between the late overpressure failure and the basemat penetration (T47) for accident sequences without the containment sprays for both low pressure (AE) and high pressure (SE, TE). scenarios.
The failure time for the late overpressurization failure mode is much longer than previously calculated for other large dry containment.1,3,'
This is as a result of the very high failure pressure for the Seabrook con tainment. As a consequence of this high containment failure pressure (medi'an pressure of 211 for wet and 187 psia for dry
- sequences) it is difficult to challenge the containment integrity by any conceivable event.
Hydrogen deflagration early in the accident sequence or later after vessel failure when steam condensation occurring as a. result of reactivation of sprays (due to regaining of ac power), or other natural heat sink .mecha-nisms,7 which can produce a deinerted atmosphere is not expected to challenge the containment integrity.
The impact of changes in the containment failure distribution discussed l in 3.2.5.4 is not significant for late failures. l Basemat Penetration Failure (S4, T47) can result in the absence of containment heat removal system (sprays) for a dry cavity. A 26-inch high curb surrounds the reactor' cavity that prevents the entry of water into the cavity unless all of the water from the RWST has been injected. The conditional probability of the basemat melt through is usually less than the late overpressurization failure, this is particularly true for Seabrook where there is a natural bed rock formation directly under the basemat foundation. Therefore, the basemat penetration failure probabilities are conservatively assigned.
No Failure (SS, M) 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.
Containment Isolation Failure (S6. 33V) is represented by an 8-inch diameter purge line. The accident sequences where the containment is either not ' iso-lated or bypassed (Event V) are assigned a conditional probability of unity to i these release categories.
1 In the SSPSA, the conditional probability for failure to isolate contain-ment (s-failure mode) is assumed to be negligibly small. This is believed to be an. optimistic assumption on the part of the applicant, because even for subatmospheric containment the 8-failure mode is expected to have a
- For cry sequences, only primary system water inventory is available in the containment. In this case, the containment atmosphere becomes superheated nnd, at failure, the temperature can exceed 700*F.
conditional failure probability ranging from 4x10 4 to about 2x10 3; there-fore, one expects the conditional probability 8-failure mode probability for a large dry containment to be somewhat higher, and perhaps approaching -10 2, An interfacing systems LOCA (V sequence) results from valve-disc rupture or disc failing open for series check valves that normally 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 SSPSA, only three possible interfacing 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 stem mount-ed limit switch and " disc failing open while indicated closed" in the other motor-operated valve in the normal RHR hot-leg suction.
For the V-sequence, the core melts early with a low RCS pressure and a dry reactor cavity at vessel melt-through. The containment sump remains dry 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 agreement with the other previous studiesl,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 uncertainties in the severe accident progression. 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 significantly more pro-nounced, 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 i
are sumarized in Table 3.10. Table 3.10 indicates that only eight of 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 containment 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), 5T and S3V dominate the source tenn frequency.
Table 3.10 Frequency of Dominant Nelease Categories (yr-1)
L Internal, Fires, .
Floods and Truck Internal and Category Crashes Seismic <0.5g Seismic >0.5g bternal
! 53 7.50E-7 3.45E-8 2.69E-7 1.05E-6 1
S5 5.64E-5 1.52E-6 1.23E-6 5.92E-5 i
S2
- 1.12E 2.40E-7 3.52E-7 S3 5.50E-5 1.10E-6 1.76E-6 5.79E-5 1
- 1.25E-5 S2Y 5.29E-6 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 I'
56V 1.80E-6 1.66E-7 3.93E-7 2.36E-6 i
)
,+.,--n,,, , _.- . , . ,---.,.n , _ ,- ,.g,.._-
Table 3.11 Contribution of Containment Response Classes to the Total Core Melt Frequency Internal, Fires, Internal Containment Floods and Truck and Class Crashes Seismic <0.59 Seismic >0.59 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 1, T
6 0.01. - - -
0.01 7 *- * * *
- 8 0.025 0.055 0.08 0.08 9-13 * * * *
- I e
a 4
4
Table 3.12 Release Category Frequency as a Fraction of Core Melt Frequency Release Internal, Fires, Internal Category Floods and Truck and Crashes Seismic <0.59 Seismic >0.59 Total Seismic External S3 <0.01 <0.01 <0.01- <0.01 <0.01 S6 0.25 <0.01 <0.01 0.01 0.26 iCf * <0.01 <0.01 <0.01 <0.01 3Ci 0.24 40.01 <0.01 0.01 0.25
- 0.03 S2V 0.05 0.08 0.08 1.
Y S3V 0.33 0.01 0.01 0.02 0.35 S4V 0.04 <0.01 <0.01 <0.01 0.04 S6V 0.01 <0.01 <0.01 <0.01 0.01 i
f 4
, , -4 8-
- 4. ACCIDENT SOURCE TERMS
~
In this chapter the approach utilized in the SSPSA to determine the frac-tion of fission products (originally in the core) that might be leaked to the outside environment will be outlined. The fission product source to the en-vironment, as calculated by this approach, will also be discussed forL each release category.
4.1 Assessment of Severe Accident Source Terms The CORRAL-II code was used in the 'SSPSA for determining fission product leakage to the environment. This code takes input from the thermal-hydraulic analysis carried out for the containment atmosphere. In addition, it needs the time-dependent emission of fission products from the damaged fuel. The fission products were assumed to be released in distinct phases as suggested in the RSS,13 namely, the Gap, Melt, and Vaporization phases. The time depen-dence of these phases is determined by the timing of core heatup, primary sys-tem failure, and start of core / concrete interactions. The methods used in the SSPSA differ from the RSS methods in the following ways:
- 1) The treatment of iodine was changed. Iodine was assumed to be in the form of cesium iodide. This was accomplished by merely using the same fraction of core inventory released for both the cesium group and the iodine group.
- 2) Leakage releases are represented by a multi-puff model,
- 3) An uncertainty analysis was carried out in which it was attempted to 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 (as shown in Table 3.8) ranging from containment bypass sequence to the no-fail sequence .
These release categories were evaluated by the applicant considering the containment failure mode, the availability of the spray system, and whether or not the cavity was wet or dry. Table 4.1 shows the point-estimate releases as -
determined by the methods outlined above. Containment failure mode S1 corre-sponds to a gross failure of the containment, resulting from a steam explo-sion, early pressure spike, or early ' hydrogen' burn. Failure mode S2 repre-sents a loss of containment function 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 represents a late overpressurization failure of the containment driven by decay heat or late hydrogen burn. Failure mode S4 represents a basemat melt-through, SS represents no containment failure and the leak rate is limit-ed to the containment design basis leak rate. Finally, failure mode S6 repre-sants sequences where the containment is failed or bypassed as part of the initiating event.
The second parameter considered in defining the source term is the avail-ability of sprays. This is determined by the plant damage states. Those l I
s.
Table 4.1 Seabrook Point-Estimate Release Categories Time of Seabrook Initiation.of Release Energy Release Release Fractions by Group Release Accident Rele.'e Duration Warnin9 Time Release Height Category Sequence (hours) (hours) (hours)* 10" cal /sec (meters) Xe 1 -2" Cs Te Ba Ru La St AEC- 1.9 0.5 0.35 < 10.0 .94 023 10 023 .24 0033 .41 9.8-5 52 AE C ' 2.6 1.0 1.9 < 10.0 10 89 2.1-5 2.1-5 4.4 2.9-6 8.8-7 8.8-8 S3 IE4 66.1 0.5 62.5 '210.0 10 .90 1.0-7 1.0-7 1.9-8 1.3 .8 3.8-9 3.8-10 55 AEC 1.9 24. 0.35 < 10.0 10. 0091 3.5-8 3.5-8 6.1 9 4.0-9 1.2-9 1.2 10 56 IEC1 4.5 4.0 4.0 < 10.0 10 90 .0036 .0036 .00067 .00044 .00013 1.3-5
^
5T AL 1.4 0.5 0.3 210.0 10. 94 .75 , .75 .39 .093 '.46 0028 57-1 7.3 9.1 6.2 < 10.0 10 .15 .092 .092 .017 .011 0034 .00034 57-2 20.3 17. 19.2 < 10.0 10 .24 093 .093 .017 012 0034 00034 57-3 29.3 1.2 28.2 < 10.0 10. .51 .12 .12 023 .015 0046
- 00046 57 Total AL 7.3 ,
2}.3 6.2 --
- 10. .90 .31 .31 .057 .038 .011 .0011 U ' AL 21.2 0.5 '26.4' 210.0' 10. .90 .122 .122 022 .015 .0044 00044
- 55. 'TEC 4.3 24. 0.6 < 10.0 10. .014 5.2-7 5.2-7 9.5-8 6.3-8 1.9-8 1.9-9 &
c-57V-1 2.2 3.5 1.9 < 10.0 10 .05 .037 .037 0069 .0045 0014 00014 57V-2 6.2 7.2 5.9 < 10.0 10 .10 .012 .072 , .0080 0079 0062 0010 52V-3 . 35.2 78.0 +34.9 < 10.0 10. .85 .20 .20 .30 022 .018 0030 52V Total AE 2.2 88.7 1.9 --
- 10. 1.0 .31 .31 .32 .0 34 025 0012 57W ^TE 81.5 0.5 76.2 210.0 10 1.0 024 .024 .030 0026 0'12 3 00039 51V AE 50.0 0.5 49.6 210.0 10. 1.0 .058 .058 .012 .0062 .0054 00091 HV-1 2.2 1.0 1.7 .35 10 .15 .11 .11 .020 .014 ~ 0041 0(X)41 S.6V-2 4.2 3.0 3.7 .33 10 .31 .14 .14 .026 .017 .0052 00051 56V-3 11.2 10.0 10.7 .24 10 .51 .18 .18 .36 017 .024 0044
$iiV Total SE1 2.2 14.0 1.7 .26 10 .97 .43 .43 40 048 .033 0053 NOIE: Exponential notation 15 showri in abbreviated form; f.e. 2.1-5 = 2.1 s 10 5
- 8ased on time of gap release escept for 56 and T6V ihere it is based on 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after accident initiation.
" Elemental lodine - not used, all fodine is treated as Cst. .
release categories with operating. spray systems are designated S1 to S6, while
. those with spray systems not operating are designated 3T to T6.
The thi rd and final parameter considered in differentiating between source terms distinguishes between wet and dry cavities. In the case of a dry cavity a vaporization release due to core / concrete interactions was assumed to '
occur. For a wet cavity the core debris was assumed to be either quenched or it was assumed that the water in the cavity would scrub the vaporization re-lease thus effectively reducing the vaporization release to zero. The release categories, which include a vaporization release, are shown in Table 3.8 and include the V sequence.
From the point of view of risk it was found5 that S2V, E $TV, and S6V 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 U and M have late overpressurization failure modes, with no spray systems operating and differ only in the treatment of the vapor-ization release phase. The containment at Seabrook is calculated 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 resulting in a puff release of approximately 0.5 hr release duration. From Table 4.1 it is seen that the H and W sequences fail at 27.2 hrs and 91.5 hrs, respectively. These failure times are several hours later than was calculated for Indian Point, Zion, and Millstone-3. The primary reason for the later failure in this case is due to the high calculate capacity of the Seabrook containment structure.
Table 4.2 compares the E S3V release parameters with similar parameters for the other three reactors analyzed in References [1], [3], and [43 Note that a fair comparison should set (0I+I) equal to (Cs-Rb), since iodine was treated as Csl in the SSPSA but not in the other studies. It is seen that I, Cs, and Ba groups for S3 are approximately half the other releases, while the Te, Ru, and La groups are low by approximately an order of magnitude. This difference is due to the later failure time in the.SSPSA (allowing more time for aerosol settling) and the absence of a vaporization release, (which dominates the release of Te, Ru, and La). A similar comparison for the TJV release indi-cates a uniform reduction of approximately an order of magnitude for all spe-cies. The reduction is entirely due to the late failure time for- this sequence.
Another importcnt consideration is the increased rate of fission product 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 containment due to changes in the containment thermal hydraulic conditions.
Release category TS is associated with early containment failure in which the containment function is compromised by increasing the leakage area in such a way that the leak rate increases from 0.1% per. day to 40% per day.
This release rate is not enough to prevent an ultimate overpressurization failure.. This release is modeled as a multi-puff release. The first puff corre'sponds to the release of fission products prior to the core debris melt-ing through the reactor pressure vessel (melt + gap). . The second puff includes
' ~
.]
Table 4.2 Late Overpressurization Failure Comparison 5
Millstone-3 7 Zion / Indian" Indian 3 Seabrook Point Study Point
~T S Sf7 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) 9.3(-2)
Cs-Rb 1.2(-1) 2.4(-2) 3.0(-1) 3.4(-1) 2.6(-1)
Te-Sb 2.2(-2) 3.0(-2) 3.0(-1) 3.8(-1) 4.4(-1)
Ba-Sr 1.5(-2) 2.6(-3) 3.0(-2) 3.'(-2) 2.5(-2)
}
Ru 4.4(-3) 2.3(-3) 2.0(-2) 2.9(-2) 2.9(-2)
La 4.4(-4) 3.9(-4) 4.0(-3) 4.9(-3) 1.0(-2)
T (release) 27.2 81.5 20 (hrs)
T (duration) 0.50 0.50 0.50 0.50 j (hrs)
Energy 300E7 300E7 540E6 15CE6 (Btu /hr)
O v y. q- m-
- g.W> q.w. - - - - - - -
- -- e a'mo, .---., - - .c.. . - .m - -- ,- em
' , h '.
the . period of vaporization release and the third puff is equivalent to an
!; . overpressurization failure ~ at the time of catastrophic containment failure.
In this model the duration of the melt release is seen to be 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />,- vapor-ization release 7.2' hours and the remaining release 78.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. The long release time for this release is not tied to the release timing, i.e., melt release and vaporization release time, but determined by the blowdown charac-teristics of the containment building. In 'rder o to determine the consequences
- from this release using the CRAC code the duration of release will have to be i limited to the maximum allowed by the code model. . Currently, this mode limits '
[ duration times 'of approximately 10 hrs. The suggested values for use in CRAC are shown in Table 4.7. Also shown in Table 4.7 is the energy of release for l the various releases. It should be noted that the high value of this param-i eter relative to the corresponding RSS value is due to a different failure mechanism in the current study. The Seabrook containment building is a rein-
~
forced concrete structure where as the RSS considered a pre-stressed concrete structure. Furthermore, the failure pressure of the Seabrook containment is
- approximately 50 psi higher- than the RSS containment failure pressure. -This will also tend to increase the energy of the released plume. ,
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 flf release is approximately half. '
the M-4 release. Without the benefit of a calculation, it is difficult to i judge whether the differences are reasonable, i
Release category S6V has binned into it an isolation failure correspond-
! ing to - an 8" diameter breach in containment and the interfacing 1.0CA (V-se-j quence). This sequence is also represented by a multi-puff release. In this f case, as in the-previous case, the. total release time is long compared to the
{ limited release times of the RSS13 consequence model.
The release fraction is compared (Table 4.3) to the M-4 release 'from the Millstone-3 study, PWR-2 from the RSS and the V-sequence from the RSSMAPIs study for Surry. 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 4 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-l tion. First, uncertainty .in timing of major events 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) Uncertainties in timing of key events, and-
- 2) Analysis methods and data.
__ _ _ . ~ _ , . _ ._ - ___ . _ . _ . . , , . _ _ _ _ _
,. ~4_.
l Table 4.3 Comparison of Releases for Failure to Isolate Containment '
and the By-Pass Sequence 5
Seabrook Millstone-3 7 RSS13* RSSMAP 15 32V' 337 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)
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 2.0 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-3.1 I
The above principles were used in the SSPSA to detennine source term mul-4
. tipliers which would give a range of fission product leakage to the. environ-ment. . A probability was associated with each source term, and for later overpressurization failure modes (ST, S3V, and STV) the following discrete probability distribution was used in the SSPSA:
! Subcategory Probability U-a .02 U-b - .08 4
U-c .30 U-d .60 This indicates, for example that there is an 8% confidence level that .U-b correctly defines the source term for the U release category.
The' results of this analysis for the overpressurization failure modes is:
Particulate Release Factor (multiplier)
, Probability U S3V S2V
.02- .22 .63 . 17
.08 .071 .22 .07
.30 .024 .065 .02
.60 .0071 -.021 .007 From this table it is seen that for the' most likely release, i.e., "d", the reduction factors of the source term are substantial .
21041 {he first two releases can be compared to releases published in BMI-3 Volume V (Surry) for the TMLB'-c and AB-c sequences. These two I
sequences correspond to late containment. failures and are both binned into U
~
and TJV 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 S3 and S3V bracket or exceed the mechanistic estimates carriedoutinBMI-2104forboththeTMLB'andABsequengs. However, for the
- less volatile species Te, Ba, .Ru, and La,' the . BMI-2104 calculated releases I
for the TMLB' sequence are higher. than all' the U and '5W releases. This dis-
" crepancy 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 U and S3V. No comparative sequence for L wit was. analyzed .in BMI-2104, i
a l
.i
Table 4.4 Comparison of AB-c and TMLB'-c (BMI-2104) to S3V and 53 Release Fractions
. Release Probability Release Category Time (hrs) Xe Cs l To Be 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) 53V-b 08 36 9.0(-1)- 5.3(-3) 5.3(-3) 6.6(-3) 5.7(-4) 5.1(-4) 8.6(-5)
$3V-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) 5-a 02 22 1.0 2.6(-2) 2.6(-2) 4.9(-3) 3.3(-M 9.7(-4) 9.7(-5) 5-b 08 28 9.0(-1) 8.5(-3) 8.5(-3) 1.6(-3) 1.l(-3) 3.lt-4) 3. l f-5) 5-c 30 34 8.0(-1) 2.9(-3) 2.9(-3) 5.3(-4) 3.6(-4) 1.l(-4) 1.l(-5)-
5-d 60 53 7.0(-1) 8. 5(-4 ) 8. 5(-4 ) 1.6(-4) 1.l(-4) 3. l (-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) 4
- e 0
0
o
. In the case of the "S6V release category a different probability distribu-tion was used. This change reflects the break location, which initiates the 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 S6V 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 S6V 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.
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 (S6V). 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 in Tables 4.6 and 4.7 can be used to estimate '
the health and economic effects (consequences) due to radioactive atmospheric releases as a result of core melt accidents in the Seabrook Station.
The resulting consequences together with the frequency of radiological releases will enable the establishment of the severe accident risk at the Seabrook site.
l l
1 l
Table 4.5 Comparison of S6V-(cum) to V-sequence (Surry)
Release Fractions Release Probability Category Xe Cs I Te Ba Ru- La 1
S6V-a .02 .97 4.3(-1) 4.3(-1)~ 4.06 (-1 ) 4.2(-2) 3.32(-2) 5.3(-3) 1 i.
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) ,
5,
i S6V-d .08 .97 5.2(,-2) 5.2(-2). 1.3(-2) 6.6(-3) 2.0(-3) 2.2(-4) 3 V (dry) -
1.0 5.52(-1) 1.99(-1) 1.2(-1) * *
- V -
1.0 1.04(-1) 3.84(-2) ~2 .5(-2) * * *
(submerged)
J
?
- Individually not reported.
s :
. Table 4.6 BNL-Suggested Source Terms Release Category Xe OI 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 SS 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 lCf 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 S3 0.90 -
0.12 0.12 0.022 0.015 4.4E-3 4.4E-4 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 -
S6V 0.97 -
0.43 0.43 0.40 0.048 0.033 5.3E-3 S6V-d 0.90 -
0.043 0.043 0.040 ~4.8E-3 3.3E-3 5.3E-4 i 1
- S6V-d release is 1/10th of the S6V values.
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 S2 2.6 1.0 <140E6 10 1.05 S3 66.0 0.5 300E6 10 63 55 1.9 '10 <140E6 10 0.35~
SE 4.5 4 <140E6 10 0.50 S1 1.4 0.5 300E7 10 0.30 TE 27 10- <140E6 10 26 TfV 35 10 <140E6 10 35 53 27 0.5 <140E6 10 26 S3V 81 0.5 300E7 10 76 S4V 50 0.5 300E7 0 49 SS 4.3 10 <140E6 10 0.30 S6V 2.5 1.0 3.7E6 10 1.0 S6V-d 2.5 1.0 3.7E6 10 1.0
- Warning time is defined as.the time after- core melt starts to the time of radiological release.
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- 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 term estimates for severe core melt accidents. ,
The containment response to severe accidents .is iudged to be an important factor in mitigating the severe accident risk. There is negligible probabil-a ity of prompt containment failure or fa.ilure to isolate. Failure during the first few hours after core melt is also unlikely and the timing of overpres-sure failure is very long compared to the Reactor Safety Study (WASH-1400).
Most core melt ace.idents would be effectively mitigated by containment spray operation. A c'.mparison of SSPSA and RSS containment failure frequencies is given in Table 5.1.
Our assessmerit 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, a
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 term study. carried out by the Accident Source Tenn Program Office (ASTP0)' of the NRC 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, i
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Table 5.1 Comparison of SSPSA and WASH-1400 l Containment Failure Frequencies l l
% of Core Melt Frequency WASH-1400 SSPSA Gross, Early Failure 34 1
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Gradual 0verpressure or Basemat Melt-through Failure 66 73 Intact Centainment --
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. 6. REFERENCES
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- 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 19R2).
- 4. " Millstone Unit 3 Probabilistic Safety Study," Northeast Utilities (August 1983).
5 B. J. Garrick, et al., "Seabrook Station Probabilistic Safety Assessment," Dickard, 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).
- 7. H. Khatib-Rahbar, et al., " Review and Evaluation of the Millstone linit 3 Probabilistic Safety Study: Containment Failure Modes, Radiological Scurce Terms and Off-Site Consequences," NUREG/CR-4143, BNL-NUREG-51907 (September 1985).
- 8. R. O. Hooten and H. Avci, " MARCH: Meltdown Accident Response Character-istf cs - Code Description and tiser's Manual," BMI-2064, NUREG/CR-1711 (1980).
- 9. J. F. Muis, et al., "CORCON-Mod 1: An Improved Model for Molten Core / Concrete Interactions," SAND 80-2415 (1981).
- 10. R. 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'?.9, 1984
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See, also, A. K. Agrawal and R. E. Hall.
- 11. W. Lycn (organizer), "RCS Pressure Boundary Heating During Severe Acci-dents," USNRC Meeting, Bethesda, Maryland (May 14,1984).
- 12. " Estimates of Early Containment loads From Core Melt Accidents," Con-tainment Loads Working Group, NilREG-1079 (Draft 1985).
- 13. " Reactor Safety Study," 11.S. Nuclear Regulatory Commission, WASH-1400, NtlREG-75/014 (October 1975). .
14 " Preliminary. Assessment of Core fielt Accidents at the Zion and Indian ;
Point Nuclear Power Plants and Strategies for Mitigating Their Effects," l NUREG-0850, Vol. 1 (November 1981). ~
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- 15. G. S. Kol b , et al . , " Reactor Safety Study Methodology Application Program: Oconee #3 PWR Plant," NUREG/CR-1659/2 of 4. '
- 16. J. A. Gieseke, et al ., "Radionuclide Release Under Specific LWR Accident Conditions," Battelle' Columbus Laboratory Reports BMI-2104 ' (July 1984, Draft).
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