ML20102C394
| ML20102C394 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 03/01/1985 |
| From: | CLEVELAND ELECTRIC ILLUMINATING CO. |
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| References | |
| NUDOCS 8503060221 | |
| Download: ML20102C394 (225) | |
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THE CLEVELAND ELECTRIC ILLUMINATING CO.
PRELIMINARY EVALUATION OF THE PERRY NUCLEAR POWER PLANT HYDROGEN CONTROL SYSTEM I
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Docket Nos. 50-440 50-441 March 1, 1985 R
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PRELIMINARY EVALUATION OF THE PERRY NUCLEAR POWER PLANT HYDROGEN CONTROL SYSTEM
1.0 INTRODUCTION
2.0 HYDROGEN CONTROL SYSTEM DESCRIPTION 2.1 I 2.2 2.3 Introduction Design Criteria Igniter Description 2.4 Igniter Locations 2.5 Power Supply and Controls 2.6 Testing 2.7 System Operation 3.0 CONTAINMENT AND DRYWELL ULTIMATE CAPACITIES I 3.1 3.2 Containment Ultimate Capacity Drywell Ultimate Capacity 4.0 CONTAINMENT ANALYSIS 4.1 Introduction 4.2 Event Scenario 4.3 Containment Response to Hydrogen .
Combustion (CLASIX-3) 5.0 DESIGN COMPARISON TO GRAND GULF 5.1 Introduction 5.2 HCS Design I 5.3 5.4 Containment Structural Capacity Containment Systems Design 5.5 Containment Response Analysis l Appendix A 5.6 Survivability of Essential Equipment
" Containment Pressure And Temperature Response To Hydrogen Combustion For Cleveland Electric I Illuminating Co. Perry Nuclear Power Plant" (OPS 38A92)
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1.0 INTRODUCTION
The Perry Nuclear Power Plant (PNPP) Combustible Gas Control System, as described in FSAR Subsection 6.2.5, is a redundant safety-grade system designed to meet the requirements previous-ly set forth in 10 C.F.R. 50.44 (i.e., prior to recent amend-ments discussed below.) It consists of two 100% capacity hy-drogen recombiners, a drywell purge system, and a backup I containment purge system. The system provides the capability to control the hydrogen which may be generated from a postu-lated design basis accident.
The accident which occurred at TMI Unit 2 resulted in the gen-eration of hydrogen beyond the limits previously specified in 10 CFR 50.44. This excessive hydrogen generation was primarily due to premature termination of the emergency core cooling sys-tem. Measures taken subsequent to the TMI-2 accident, (e.g.
compliance with NUREG-0737 requirements) along with the inher-I ent resistance of the BWR 6/ Mark III plant to events which could result in a degraded core, effectively precludes the need for other systems to prevent or mitigate the consequences of the generation of large amounts of hydrogen.
The PNPP BWR 6/ Mark III design features which provide inherent resistance to degraded core events and protection against plant damage and release of radioactivity in excess of 10 CFR Part 100 limits are:
- a. Numerous automatic high and low pressure pumps which provide makeup water to the reactor vessel.
- b. Rapid depressurization capability via the Automatic Depressurization System.
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d.
Natural circulation internal to the reactor vessel.
Two above core spray systems for core cooling.
- e. Direct reactor vessel water level measurement.
The capability to vent noncondensible gases from the I f.
reactor vessel.
- g. A large suppression pool heat sink for decay heat l h.
removal.
Suppression pool scrubbing of fission products.
- i. Secondary containment providing an additional barrier to radioactive releases.
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I Recently, the Nuclear Regulatory Commission amended the hydre gen control requirements of 10 C.F.R. Part 50.44 (50 Federal Register 3498, January 25, 1985) to require improved hydrogen I control systems for Mark III containments to handle large amounts of hydrogen during and following an accident. The new rule requires that prior to exceeding 5% reactor power, a licensee "shall provide its nuclear reactor with a hydrogen control system justified by a suitable program of experiment and analysis. The hydrogen control system must be capable of I handling without loss of containment structural integrity an amount of hydrogen equivalent to that generated from a metal-water reaction involving 75% of the fuel cladding sur-rounding the active fuel region." (Section 50.44(c)(3)(iv)(A)}.
" Completed final analyses are not necessary.for a staff deter-mination that a plant is safe to operate at full power provided that prior to such operation an applicant has provided a pre-liminary analysis which the staff has determined provides a satisfactory basis for a decision to support interim operation at full power until the final analysis has been completed."
(Section 50.44(c)(3)(vii)(B)]
The Cleveland Electric Illuminating Co. (CEI) has evaluated a number of possible system concepts for controlling the genera-I tion of large amounts of hydrogen. The technical criteria used to assess these various options considered the mitigation effectiveness, consequences of intended or inadvertent opera-I tion, reliability, testability, availability of design and equipment, and impact on the public health and safety (if any).
CEI chose a hydrogen combustion system as the most viable con-cept for PNPP. A hydrogen ignition system has been designed I and is being installed at PNPP. The system will be tested and operable prior to exceeding 5% reactor power.
This document provides a preliminary evaluation which meets and exceeds the preliminary analysis requirements of the new hydro-gen rule. A detailed description of the PNPP Hydrogen Control I System (HCS) is provided in Section 2.0 of the document. A significant amount of plant specific analysis has been con-ducted to support the preliminary evaluation. This includes analyses of the containment pressure capacity, discussed in Section 3.0 of this document, and analyses of the containment pressure and temperature response to hydrogen combustion, dis-cussed in Section 4.0. Section 5.0 of this document provides a comparison of the PNPP HCS key design features and supporting analyses to those of the Mississippi Power & Light (MP&L) Grand Gulf Nuclear Station (GGNS), which the NRC has licensed for I full power operation on an interim basis. This comparison es-tablishes the similarity of systems and provides additional basis for a decision by the NRC to support full power operation for PNPP.
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2.0 HYDROGEN CONTROL SYSTEM DESCRIPTION I
2.1 INTRODUCTION
The Hydrogen Control System at PNPP is an ignition system, which consists of igniter assemblies distributed throughout the drywell, wetwell and upper containment regions of the plant.
I It is designed to handle, without loss of containment structur-al integrity, an amount of hydrogen equivalent to that gener-ated from a metal-water reaction involving up to 75% of the fuel cladding surrounding the active fuel region. This is ac-I complished by burning hydrogen at low concentrations, thereby maintaining the concentration of hydrogen below the detonable limits and preventing containment overpressurization failure.
The potential for significant pocketing of hydrogen will be precluded by:
- a. utilization of distributed ignition sources,
- b. operation of containment sprays,
- c. mixing caused by turbulance resulting from lo-calized burns.
The hydrogen ignition system is designed with suitable redun-dancy to assure that no single active component failure, including power supply failures, will prevent functioning of I the system. The system is designed as a safety grade system, and is capable of operating for the duration of the hydrogen generation event.
I 2.2 EQUIPMENT DESIGN CRITERIA The Hydrogen Control System igniter assemblies are classified as electrical safety Class lE and seismic Category I. This equipment is designed, manufactured, tested, and certified in I accordance with the following standards:
- 1. American National Standards Institute (ANSI)
N45.2.2 - 1972 " Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants (during the construction phase)."
- 2. Institute of Electrical and Electronic Engineers (IEEE) Standards:
- a. IEEE-308, (1974) " Standard Criteria for Class IE Power Systems for Nuclear Power Generation Stations."
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- b. IEEE-323, (1974) " Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations."
- c. IEEE-344, (1975) " Recommended Practices for Seismic Qualification of Class lE Equipment for Nuclear Power Generating Stations."
- d. IEEE-383, (1974) " Standard for Type Test of lE Electric Cables, Field Splices, and Con-I nections for Nuclear Power Generating Sta-tions."
- 3. U.S. Nuclear Regulatory Commission Regulatory Guide NUREG-0588, Revision I, Category 1, "In-terim Staff Position on Environmental Qualifica-tion of Safety-Related Electrical Equipment."
- 4. American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code:
- a.Section II, " Material Specifications," 1980 Edition through Summer 1982 Addenda.
- b.Section IX, " Welding and Brazing Qualifica-tions," 1980 Edition through Summer 1982 Addenda.
2.3 IGNITER DESCRIPTION The igniter assemblies used in the Hydrogen Control System are I divided into two components:
- a. the igniter enclosure which partially encloses I the igniter and contains the terminal block, transformer, and associated electrical wiring, and
- b. the junction box which contains the cable termi-nation.
The assembly is depicted in Figure 2.3-1. Spray shields are provided for igniter assembly protection in areas where the ig-niter may be exposed to containment sprays.
The igniter enclosure, junction box, and spray shield are con-structed of stainless steel. The enclosure is 1/8 inch thick, I the junction box is 14 gauge. Gasketing material and sealant 3 is provided to ensure leak-tightness of the igniter enclosure and junction box.
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The entire igniter assembly used at PNPP is identical to that used at the Grand Gulf Nuclear Station. The igniter, shown on Figure 2.3-2, is a General Motors AC Division, Model 7G glow plug. The transformer is a 0.2 KVA Dongan Model 52-20-472,120+
10% VAC, 60 Hz primary with multiple secondary taps. The ig-niter assembly is manufactured by Power Systems Division of Morrison Knudsen.
2.4 IGNITER LOCATIONS Igniter assembly design locations have been finalized and are I given on Table 2.4-1. As-built igniter locations will be es-tablished during installation depending on availability of sup-ports and interferences in the areas identified. The igniter locations are based on the following criteria:
- a. Hydrogen can be released directly to the con-tainment atmosphere through the safety relief valves which exhaust to the suppression pool.
Igniter assemblies are located in a ring at elevation 619'-6", which is above the suppres-I sion pool, directly below the HCU floor. This assures combustion of the hydrogen release close to the pool surface (elevation 593').
I b. In open areas of the containment up to the refueling floor and for all areas of the I drywell, except for the post-LOCA reflood area, the igniter assemblies were located using the following criteria:
- 1. Assuming only one Engineered Safety Feature (ESF) power division is functional follow-ing an accident, a distance of 60 feet ex-ists between operable igniters. In some cases, the distance may be up to 70' feet if supports are not available or interfer-ences exist in the areas identified.
- 2. Assuming both ESF power divisions are func-tional following an accident, a distance of 30 feet exists between operable igniters.
In some cases, the distance may be up to 35' if supports are not available or inter-I c.
ferences exist in the areas identified.
For enclosed containment areas which could accu-mulate hydrogen, two igniters are located in each room. A separate ESF power division sup-plies each igniter.
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- d. Hydrogen can be released directly to the drywell atmosphere via a small pipe break in the drywell. Igniter assemblies are located throughout the drywell.
e.. Igniter locations in the drywell take full ad-I vantage of existing steel as protection against jet impingement loads and are spaced so that one jet cannot impair more than one igniter.
Based on the above criteria, 102 locations in the containment and drywell will have igniter assemblies. The number and ar-rangement of igniter assemblies are similar to those at the Grand Gulf Nuclear Station. Figures 2.4-1 through 2.4-11 show the location of PNPP igniter assemblies in the containment as well as their relative location to major equipment or struc-I tures. Figures 2.4-12 through 2.4-16 show PNPP cross-sectional containment flow areas at different elevations within contain-ment.
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2.5 POWER SUPPLY AND CONTROLS The hydrogen igniters are powered from 120VAC, 60 Hz, Class lE power distribution panels M56-P003 through P008. These power panels receive their power from Class lE 480V motor control centers (EFlA08 or EFlCO8) through 15 KVA transformers, rated 480-208/120VAC, 60 Hz, 3-phases with grounded neutrals, and a I fuse panel (M56-P001 or P002). The fuse panel consists of a 40 amp and 45 amp fuse in series for each line to the 120 volt distribution power panels. Each transformer is fed from a I Class lE MCC breaker from a Class lE bus which is capable of being powered from one of the Standby Diesel Generators.
The 102 igniters are divided into six groups of approximately equal number, three groups in Division I and three groups in Division II. Each group is powered from a separate distri-I bution power panel. The power panel disconnect switches are provided for maintenance and are normally closed so the ig-niters can be energized by operating control room handswitches.
The igniters are manually energized by means of two handswitches located in the control room on Panel H13-P800.
One for the three Division I groups (M56-S1) and one for the I three Division II groups (M56-S2). The switch positions are ON-OFF with red-green indication lights. Input is provided to the hydrogen control system out-of-service annunciator at panel H13-P800 on loss of control or motive power.
Table 2.4-1 lists the divisional power supply to each igniter and Figure 2.5-1 contains a simplified electrical schematic for I the Hydrogen Control System.
l 2.6 TESTING 1
2.6.1 Preoperational The Hydrogen Control System will be preoperationally tested to I ensure correct functioning of all controls, instrumentation, wiring, and the transformers and igniters. The test will in-
! clude energizing one of the two divisions from the control room and verifying that all igniters powered from the associated panel are functional. Identical procedures will be followed for the remaining igniters powered off of the other division.
Functional testing of the system will include verification of I the following:
1 E 1. Surface temperature of each igniter is operating l 5 at or above 1700"F with 120 VAC+10%, 60 Hz ap-plied to the igniter assembly. _
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- 2. The 480-208/120 volt transformers (M56-S201 and S202) are capable of providing satisfactory sec-ondary voltages of 120112 VAC and of meeting the I 3.
minimum load requirement of 15 KVA.
All hydrogen igniter transformers are capable of providing satisfactory hydrogen igniter voltages of 12.011.2 VAC.
2.6.2 Surveillance The HCS surveillance requirements will be included in the PNPP Technical Specifications.
2.6.3 Qualification The qualification of the hydrogen igniter assembly is in accor-dance with the PNPP equipment qualification program described I in FSAR sections 3.10 and 3.11. The hydrogen igniter qualifi-cation program meets the requirements of the following docu-ments:
o IEEE Std. 323-1974, "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations" (including the November 21, 1975 Supplement) and USNRC Regulatory Guide 1.89.
o IEEE Std. 344-1975, "IEEE Recommended Practices for Seismic Qualification of Class lE Equipment for Nuclear Power Generating Stations" and USNRC I Regulatory Guide 1.100.
IEEE Std. 381-1977, "IEEE Standard for Type Tests o
of Class lE Modules Used in Nuclear Power Generating Stations".
I o IEEE Std. 627-1980, "IEEE Standard for Design Qualification of Safety Equipment Used in Nuclear Power Generating Stations".
o USNRC NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electric Equipment".
o 10 C.F.R. Part 50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants".
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o ANSI /ASME NQA-1-1983, Quality Assurance Program Requirements for Nuclear Power Plants, o ANSI /ASME N45.2-1977, Quality Assurance Program Requirements for Nuclear Facilities.
The qualification program used to meet these requirements is described below. The program includes testing to simulate aging, followed by radiation exposure, a seismic test and acci-dent conditions tests. The test sequence is presented below:
Inspection of Equipment '
Baseline Functional Test Thermal Aging )
Post-Thermal Functional Test ]
Radiation Exposure Post-Radiation Functional Test Wear Aging Test ,
Post-Wear Functional Test '
I Seismic Test Post-Seismic Functional Test Accident Conditions Tests LOCA Test Post LOCA Functional Test Low Pressure Transient Test Post LPT Functional Test Submergence Test Post Submergence Functional Test I After thermal aging to a 40 year design life, the igniters are irradiated to achieve accident condition neutron and gamma in-tegrated doses. They are then subjected to vibration and seismic tests. The igniters are energized during these tests.
Each component of the hydrogen igniter, including subvendor subcomponents, is designed to withstand the maximum accelera-tion created by the appropriate load combinations provided in FSAR section 3.10.
The igniter assembly is tested in accordance with a LOCA envi-ronmental test profile which meets or exceeds the PNPP Environ-I mental conditions specified in FSAR Section 3.11 Tables. The test sample is operated during the LOCA test to verify op-erability under the actual environmental conditions expected to occur in service including sprays. Functional tests are per-formed following the LOCA testing.
Following the LOCA test, the test sample is placed inside a smaller pressure vessel to simulate the negative pressure tran-sient postulated to occur as part of the accident conditions.
The pressure is reduced from 9 psig to -14 psig at a rate of 20 psi /sec. The pressure remains at -14 psig for approximately two seconds after which time it will return to atmospheric I
I pressure at a rate of 4.5 psi /sec. The igniters are not op-erating during this test. An inspection is performed to deter-mine if the pressure transient test caused any deformation of the hydrogen igniters. A functional test is performed after this portien of the test.
The test sample is subjected to a water submergence test fol-lowing the pressure transient test. The igniter is arranged in a test fixture to allow the igniter to be submerged rapidly.
The igniter is submerged, while operating, for a period of ap-proximately five seconds. The igniter is removed and a com-plete functional test performed.
2.7 SYSTEM OPERATION The Hydrogen Control System is placed in service in accordance with the generic emergency procedure guidelines when the reac-tor water level reaches the top of the active fuel. The ig-niters are energized by two ON-OFF handswitches (M56-51 and M56-S2) located in the control room on panel H13-P800.
I Red-green indication lights for each handswitch are provided.
There are no interlocks associated with the Hydrogen Control System.
After manual initiation, the igniters are powered continuously for up to seven days. The system is manually de-energized by the operator turning both handswitches (M56-S1 and M56-S2) to I "OFF" when the hydrogen generation event has passed.
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FIGURE 2.3-1 GErfRAL ASSEMBLY HYDROGEN IGNITER
m m m m m m M M M M M M M m m m m m M 3/8 HEX WRENCll Fl.ATS CONDUCTOR SilEATil -[ ~
HEATER ELEMENT (INCONEl. 601) (Il0 SKINS 875) i y f l , , ff --
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ICHITION TERitINAL f
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I B TABLE 2.4-1 HYDROGEN IGNITER LOCATIONS ESF DIMENSION TO
. POWER CENTERLINE IGNITER # DIVISIONELEVATION MIMUTH OF CONTAINMENT l IM56-001 1 613'4" 3550 49'0" IM56-002 2 613'4" 50 51'0" IM56-003 2 619'6" 630 51'8" IM56-004 1 619'6" 890 52'0" IM56-005 664'0" I
1 340 57'0" IM56-006 2 689'0" 340 52'0" IM56-008 1 629'1M" 11058' 36'-6" IM56-009 2 637'0" 4105' 36'6" IM56-010 1 636'3M" 900 36'-6" IM56-011 2 636'7" 1370 36'-6" IM56-012 1 632'3" 1800 36'-6" I IM56-013 IM56-014 IM56-015 2-1 2
631'5" 636'10" 630'9M" 2210 2730 3220 36'-6" 36'-6" 36'-6" IM56-016 2 660'0" 00 31'6" IM56-017 1 659'8" 570 29'6" IM56-018 1 659'8" 1140 30'-0" IM56-019 l
l I IM56-020 IM56-021 2
1 2
659'8" 659'8" 660'0" 1720 2250 2800 30'-0" 28'0" 30'-0" IM56-022 1 660'0" 3170 31'0" I IM56-023 IM56-024 1
2 619'6" 619'6" 340 1180 52'0" 31'8" IM56-025 1 619'6" 1520 51'0" I IM56-026 IM56-027 IM56-023 2
1 2
619'6" 619'6" 619'6" 1860 2210 52'0" 51'8" 2550 51'4"
- 3 IM56-029 - 1 619'9" 2890 52'0"
' 5 IM56-030 2 619'0" 322030' 51'11" IM56-031 2 638'0" 358030' 41'-6" IM56-032 2 640'0" 1550 46'0" IM56-033 1 640'0" 186030' 46'0" l IM56-034 1 640'-0" 3240 53'6" IM56-035 2 640'4 3/4" 610 51'6" IM56-036 1 640'54" 1180 51'6" IM56-037 2 640'5" 2270 46'0" IM56-038 1 639'4 M" 260030' 54'0" IM56-039 1 651'1" 2860-30' 41'-6" IM56-040 1 647'4" 20 41'6" IM56-041 1 650'6 3/4" 410 50'-6" I IM56-042 IM56-043 2
1 650'6" 651'-0" 870 1010 49'0" 49'-0" l
I TABLE 2.4-1 (Ccntinued)
HYDROGEN IGNITER LOCATIONS ESF DIMENSION TO '
POWER CENTERLINE IGNITER # DIVISION ELEVATION AZIMUTH OF CONTAINMENT IM56-044 1 660'0" 86030" 44'6" IM56-045 2 660'6" 950 48'-6" I IM56-046 IM56-047 IM56-048 2
1 2
664'0" 665'-0" 662'6" 540 1140 1470 31'0" 52'0" 53'-0" I IM56-049 IM56-050 IM56-051 1
2 1
662'7 3/4" 664'7" 661'6" 2180 2510 2890 51'0" 49'6" 50' IM56-052 2 661'6" 3240 49'6" IM56-053 1 669'-6" 00 54'6" IM56-054 2 684'-9" 3550 52'-6" IM56-055 1 686'-0" 750 48'-0" I IM56-056 IM56-057 IM56-058 2
2 1
686'-0" 686'-0" 686'-0" 850 950 47'-0" 47'-0" 1050 48'-0" IM56-059 1 686'-0" 750 35'-0" I IM56-060 2 686'-0" 1050 35'-0" IM56-061 1 689'-6" 450 48'-0" IM56-062 2 689'-06" 1330-15' 41'-0" IM56-063 1 689'-6" 2290 48'-0" IM56-064 2 689'-6" 2520 43'-6" IM56-065 1 689'-6" 2890 43'-0" I IM56-066 IM56-067 2
1 689'-6" 715'-6" 3100 3580-51' 48'-6" 58'-9" IM56-068 2 715'-6" 270-8' 58'-9" IM56-069 ~1 715'-6" 610-47' 58'-9 "
IM56-070 2 715'-6" 870-32' 58'-9" IM56-071 1 715'-6" 1190-27' 58'-9" IM56-072 2 715'-6" 1500-33' 58'-9 "
IM56-073 1 715'-6" 1780-46' 58'-9" IM56-074 2 713'-6" 2090-27' 58'-9" IM56-075 1 715'-6" 2400-35' 58'-9" IM56-076 2 715'-6" 2730-9' 58'-9" IM56-077 1 715'-6" 3000-26' 58'-9" IM56-078 2 715'-6" 3310-38' 58'-9" i IM56-079 1 745'-6" 3580-48" 48'-0" IM56-080 2 745'-6" 340 48'-0" IM56-081 1 745'-6" 720 48'-0 "
I IM56-082 IM56-083 2
1 745'-6" 745'-6" 1020 1430 48'-0" 48'-0" IM56-084 2 745'-6" 1800 4 8'-0" I IM56-085 IM56-086 IM56-087 1
2 745'-6" 745'-6" 2160 2520 48'-0" 48'-0" 1 745'-6" 2870 48'-0" l
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I TABLE 2.4-1 (Continued)
HYDROGEN 'GNITER LOCATIONS ESF DIMENSION TO POWER CEtiTERLINE IGNITER # DIVISIONELEVATION AZIMUTH OF CONTAINMENT II IM56-088 745'-6" 3240 2 48'-0" IM56-089 2 737*.on ao l'_oa l IM56-090 2 757'-0" 1800 l'-0" 1
IM56-091 1 645'7" 1680 60'0" IM56-092 2 645'-0" 1720 58'-0" I IM56-093 IM56-094 IM56-095 1
2 1
613'-4" 612'5" 612'6" 70 12030:
44'-0" 42'8" 3430-30' 42'6" I IM56-096 IM56-097 IM56-098 2
2 1
612'3" 638'8" 658'6" 3500-30' 2890 3420 43'6" 49'6d 53'-0" I
E IM56-099 2 685'-6" 170 50'6" g IM56-100 2 686'-0" 750 25'-0" IM56-101 1 686'-0" 1050 25'-0" IM56-102 1 670'-0" 3500 13'0" IM56-103 2 670'-0" 40 13'0"
! NOTE: 1 M56-007 not used.
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PNPP UNIT I IGNITER LOCATION I REACTOR BUILDING PLAN ELEV. 593'-6"
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Figure 2.4-2 I
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Figure 2.4-3 - I
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I Figure 2.4-7 I IM56-067 THRU IM56-078 AT ELEV. 715'-6" IM56-079 THRU IM56-088 AT ELEV. 745'.6" IM56-089 THRU IM56-090 AT ELEV. 757' 0" O' 4 POWER SUPPLY'
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l d'wt I w I PLAN VIEW "B" l g "L- 617'-o" to 649'-o-E
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Figure 2.4-12 , l LEGEND
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l 2 114.7' LEVEL 4 l 180 2778 f t2 CROSS-SECTIONAL CONTAINMENT FLOW AREA I I PtfP UNIT I REACTOR BUILDING CROSS-SECTION l PLAN ABOVE ELEV. 68 9 ' -6 " I I
_ Figure 2.4-13 ; LEGEND I . _ ., .. ,. _. .._, . _ . - . - (EL 574 807 oo.n . c a r.<.
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- - 89.8' to 270 ".#$$2 25:k- 1I4.7'
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I i x .<. s l Floor at 77.2'
' 180 89.8' LEVEL
= .6 =
I CROSS-SECTIONAL CONTAINMENT FLOW AREA 2534 fL2 I
- a I PNPP UNIT I REACTOR BUILDING CROSS-SECTION PLAN ABOVE ELEV. 664'-7" l
l
Figure 2.4-14 I LEGENO NO*[ETeY.=~r.U N (EL 574*.107 osaa ... ha'. O Floor at 77.3' l e r , ,- m- - / g 67 2 ff4.7'
+ + 4' I ,
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I - 67.2' LEVEL 1 I CROSS-SECTIONAL CONTAINMENT FLOW AREA 3070 ft 2 PbFP UNIT I REACTOR BUILDING CROSS-SECTION PLAN ABOVE ELEV. 642'-0a
m Figure 2.4-15 , LEGEND
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I 3.0 CONTAINMENT AND DRYWELL ULTIMATE CAPACITIES I 3.1 CONTAINMENT ULTIMATE CAPACITY The ultimate structural capacity analysis of positive internal pressure for the PNPP Mark III containment has been evaluated. The results were transmitted to the NRC in letters dated January 25, 1982 (D.R. Davidson to R.L. Tedesco) and February I 11, 1985 (M.R. Edelman to B.J. Youngblood). Local regions of the containment vessel, equipment hatch, personnel air locks, and the main steam penetrations were evaluated for static loads. The actual material strengths of ASME-SA-516, Grade 70 steel were used in the analysis to determine the mean, lower bound and upper bound values of the material yield strength and ultimate strength. Based on these material properties, the ca-pacity of the general shell to resist statically applied pres-sure was determined to be 78 psig lower bound streroth and 94 I psig mean value strength. The limiting region of t'.c contain-ment shell for the analysis was found to be the dome knuckle. The maximum allowable pressure to meet the ASME Service Level C I limits was determined to be 50 psig for the most limiting con-tainment penetration. However, use of ASME Service Level D limits (defined in the ASME Code as " limits which are permitted I for combinations of conditions associated with extremely low probability postulated events") is a more realistic evaluation of the containment pressure capability, considering the nature E and probability of the hydrogen generation event. Utilizing 5 Service Level D stress limits, the maximum allowable pressure for the most limiting containment penetration was determined to be 56 psig. PNPP Safety Evaluation Report, Supplement 1, (NUREG-0887) Sec-tion 3.8.2 discussed the results of the containment ultimate I capacity analysis. The SER noted that the dome knuckle area contols the ultimate capacity at the containment vessel which starts to yield at 68 psig. Containment shell pressure capaci-ty can be increased to 78 psig, the pressure at which hoop I buckling occurs in the knuckle region, since yielding occurs at one point along the meridian at 68 psig. However, as previous-ly discussed, the most limiting penetration establishes the ul-timate capacity value for the containment. Previous analyses performed by the Hydrogen Control Owners Group (HCOG) Mark III member utilities have demonstrated that significant margins exist between the containment ultimate pos-itive and negative pressure capacity and the positive and nega-tive pressures postulated as a result of hydrogen combustion. I At the Grand Gulf Nuclear Station (GGNS), the ultimate capacity versus design levels are 56 psig versus 15 psig for containment I I positive pressure and the capability for containment negative pressure has been established at -10 psid versus the -3 psid design value. The PNPP margin, i.e. ultimate capacity versus design values, is similar to that of GGNS (56 psig versus 15 psig for PNPP containment positive pressure). Similarly, actual ultimate ca-pacity over the 0.8 psid design negative capacity at PNPP can be expected. Further negative pressure capability is provided in the PNPP design, which includes two 24-inch nominal diameter vacuum relief lines to assure that the negative pressure inside containment does not exceed the design value of -0.8 psid. Two additional 24-inch lines are provided for redundancy. The vac-uum breaker check valves begin to open under a negative pres-sure differential of 0.1 psid and they become fully effective in limiting the negative containment pressure within about one-half second. In the design of the containment vacuum relief capacity, two I limiting initiating events were considered for the vacuum re-lief line sizing: (1) inadvertent spray actuation following a 6 inch RWCU line break, and (2)-inadvertent spray actuation during normal plant operations. In this design basis analysis I (see FSAR Section 6.2.1.1.4.2), the following conservative as-sumptions were made to maximize the rate of cooldown due to the evaporative cooling process for case (2), which results in the lowest containment pressure: .
- a. Spray efficiency is 100%.
- b. All of the spray water entering the containment is immediately vaporized and forms a homogeneous mixture with the containment atmosphere.
- c. No heat is transferred into the containment at-mosphere from the structures during the tran-sient.
- d. Maximum temperature in containment during normal operation - 105 F.
- e. Minimum relative humidity in containment during normal operation -30 percent.
I f. Minimum spray water temperature - 60 F.
- g. Spray system flow rate - 10,500 gpm Taking credit for only two of t*c four vacuum relief lines, and using the above conservative 7.h.mptions, resulted in a maximum I negative pressure of -0.72 pt ; ' ar a 10% margin from the de-sign negative pressure va'u ' ).8 psid.
lI i I
I The margin is substantially larger when credit is taken for all four vacuum relief lines and when the ultimate (rather than the design) containment negative pressure capacity is considered. I This actual negative pressure capability at PNPP bounds the negative pressure resulting from the hydrogen burn (oxygen de-pletion), and the subsequent cooling of the containment atmo-sphere. Therefore, connidering the ultimate capacity margins that exist I for structures and the redundant containment vacuum breaker ca-pacity design, an additional analysis to calculate the ultimate negative pressure capacity of the PNPP containment is not war-I ranted. 3.2 DRYWELL ULTIMATE CAPACITY Previous analyses performed by the HCOG member utilities have also shown that significant margins exist between ultimate and I design capacity of the drywell for differential pressures (both positive and negative) resulting from hydrogen combustion. A PNPP plant-specific ultimate capacity analysis of the drywell is not warranted, based on the following conclusions based on I comparisons of PNPP and GGNS, which show, A) the drywell de-signs show similarity in structural details; and B) the CLASIX-3 containment response analyses show similarity in dif-ferential pressures resulting from hydrogen combustion for both plants. I A) The PNPP drywell is designed as a reinforced concrete structure. The primary drywell structure consists of four major components:
- 1. A flat, circular reinforced concrete foundation,
- 2. A right, vertical cylinder. The cylindrical wall is 83'x 0" outside diameter, 85'-9" high, and 5'-0" thick. The lower 26'-2" of the drywell is the vent region, composed of steel I
I and concrete composite construction. This re-gion consists of two concentric cylinders, with the annulus between the cylinders stiffened ver-3 tically by radial steel plates and filled with l g 5000 psi concrete. The upper drywell region is designed as a reinforced concrete cylinder con-nected to the lower vent region by cadwelding all vertical and diagonal rebars to the ring girder. E This upper wall is heavily reinforced with No. E 18 vertical and hoop rebars and No. 14 diagonal l rebars. On the outside face of the cylinder, I lI L
I additional No. 11 vertical rebars are provided. The upper drywell wall is integrally connected to the 4'-0" thick drywell top slab.
- 3. A flat, horizontal, circular, reinforced con-crete drywell top slab. The top slab contains a I central circular opening of 31'-11.5" diameter which is closed by the drywell head.
I 4. The 14'-9.25" deep steel ellipsoidal drywell head, which forms part of the drywell pressure retention boundary. The general arrangement and design details of the PNPP drywell structure are consistent with those previously evaluated for the GGNS. The primary drywell structure of the GGNS drywell consists of four major components:
- 1. A flat, circular reinforced concrete foundation.
- 2. A right, vertical cylinder. The cylinder wall is 75'-0" outside diameter, 91'-6" high, and 5'-0" thick. The lower 24'-10" portion of the I wall, i.e., the vent region, is of heavily rein-forced, concrete composite construction. This lower region has two stiffened steel, concen-I ,
tric, cylindrical surface plates. The annulus between the surface plates is stiffened by ver-tical, radial plates and is filled with con-crete. The upper wall is designed as a rein- ,I forced concrete cylinder which is supported by the steel, lower wall section and internal con-crete. The lower steel section is connected in-I tegrally with the upper wall vertical and diago-nal reinforcement.
- c. A flat, horizontal, circular, reinforced con-crete drywell roof slab, containing a central circular opening of about 32 feet. This opening is closed by the drywell head.
- d. A steel ellipsoidal drywell head, approximately 15'-6" deep, which forms part of the drywell I pressure retention boundary.
The structural design aspects of the GGNS and PNPP drywells are I functionally similar. Additionally, the drywell positive and negative design pressures for PNPP and GGNS are consistent in all material respects. See Section 5.4 of this report for a comparison of these values. I I
I B) The base case PNPP CLASIX-3 analyses and the CLASIX-3 sen-sitivity studies performed for GGNS provide indications of the anticipated peak positive and negative drywell pressures re-I sulting from hydrogen combustion. The assumptions and input parameters used for the PNPP analysis are similar to those used The key PNPP assumptions and for the GGNS sensitivity studies. input parameters (such as burn parameters) were the same as the corresponding GGNS cases. As discussed in Section 5.5 of this report, the results of both the PNPP and GGNS SORV and drywell I break analyses are similar. This is expected due to the simi-larity in containment design (i.e. drywell, wetwell and con-tainment volumes and heat sinks) and containment systems design (i.e. containment spray flow rates). The differential pressures (drywell minus containment) calcu-lated from the CLASIX-3 studies for GGNS, which are comparable I to those for PNPP, were approximately +9 psid and -18 psid. These differential pressures are significantly lower than the ultimate capacity of the GGNS drywell which was determined to I be +67 psid from the drywell to containment. The negative pressure capability is higher than the positive pressure capa-bility. The drywell head is capable of withstanding -89 psid. Differential pressures range from +7 psid to -11 psid for the I PNPP drywell break analysis. Given the similar drywell struc-tural designs at GGNS and PNPP, similar margins between these calculated differential pressures and the positive and negative I drywell capability are anticipated to exist for PNPP. In summary, GGNS and PNPP are similar in arrangement and design I details, and in the containment /drywell pressure and tempera-ture response to hydrogen combustion. Thus, the pressure capa-bility of the PNPP drywell structure is expected to be compara-Also, the Mark III design results in a I ble to GGNS values. pressure capacity on the order of 2-3 times that required to withstand the maximum drywell pressure differentials resulting from hydrogen combustion. For these reasons, substantial mar-gin and capability above required capacities are expected for the PNPP drywell with respect to its ability to withstand posi-tive and negative pressures associated with hydrogen burn events. I I I I
I I 4.0 CONTAINMENT ANALYSIS
4.1 INTRODUCTION
The Hydrogen Control System is designed to burn hydrogen in I small concentrations, preventing large concentrations of hydro-gen from accumulating which might ignite and threaten contain-ment integrity. As indicated in section 2.0, there are 102 ig-I niters distributed throughout the drywell and containment, which will burn hydrogen in small concentrations and prevent pocketing. CEI has conducted a preliminary evaluation using the CLASIX-3 computer code. Two analyses were performed for PNPP to inves-tigate the containment temperature and pressure response to I postulated degraded core events with deflagration burning. merous risk assessment studies have shown that transient-initiated events, as compared to accident-initiated Nu-I events, are the most probable in terms of core melt frequency. For transient-initiated events which result in a postulated re-coverable degraded core, the hydrogen release is directly into the suppression pool through a stuck open relief valve (SORV). I This event has been chosen.as the base case for the preliminary evaluation of hydrogen combustion. In order to evaluate the effect of hydrogen released directly to the drywell, the less I probable small line break in the drywell (DWB) is also evalu-ated. I The detailed report " Containment Pressure and Temperature Re-sponse to Hydrogen Combustion for Cleveland Electric Illuminating Perry Nuclear Power Plant," OPS-38A92, is attached I as Appendix A. The report includes a description of the sce-narios considered, the input assumptions, and the results, including the pressure and temperature response. This section discusses the OPS-38A92 analysis and the results. I Determination and evaluation of the containment thermal envi-ronment due to hydrogen combustion for higher hydrogen release I rates associated with diffusion burning vill be addressed in the final analysis. 4.2 EVENT SCENARIO To evaluate the role of igniters in accident mitigation, CEI I has undertaken a preliminary analytical effort to determine the effectiveness of the igniter system in reducing the threat to containment integrity caused by the combustion of hydrogen gen-I erated following postulated degraded-core accidents. Addition-al analysis and testing will continue as part of the long-term program to support the final analysis of the HCS. I I
f The preliminary evaluation of the HCS is based on the analyses of two degraded-core accident scenarios: (1) a small break { loss-of-coolant accident (LOCA) with temporary failure of emer-gency core cooling (ECC) injection, and (2) a transient with a stuck-open relief valve (SORV) accompanied by a failure of the { ECC system. The SORV was chosen as the base case recoverable degraded-core ( event, because of risk studies showing that it has a higher core melt frequency than the LOCA event. The small break LOCA was included for evaluation in order to consider the potential ( consequence of hydrogen release directly to the drywell. In order to perform analyses of the containment atmosphere [ pressure and temperature response resulting from a ( degraded-core accident, the releases from the reactor coolant primary system, including steam and hydrogen release rates, must be established. The PNPP containment response analysis ( was based on the reactor coolant system response and releases using results from the MARCH computer code. MARCH models the release of hydrogen and steam from the open-f ings in the primary system appropriate for the scenario (SORV or small line break). The two sequences. evaluated in the PNPP preliminary evaluation used identical mass and energy releases. { For the small break LOCA, hydrogen and steam enter the drywell, as well as the suppression pool through the safety relief valves. For the SORV event, hydrogen and steam are directly ( introduced into the suppression pool through the safety relief valves. This combination of releases is representative of a variety of ( recoverable degraded core situations in which hydrogen is a factor. As in the GGNS analyses, quenching and recovery were not mechanistically calculated since mechanistic scenarios can- { not produce recoverable events with 75% metal water reaction. The hydrogen reaction was terminated when 75% oxidation of the cladding was reached consistent with the new hydrogen rule. ( The fraction of fuel clad oxidized in the calculation (75%) ex-ceeds that estimated to have occurred in the TMI-2 accident (45-50%). { The MARCH steam, hydrogen, and fission product energy releases calculated for GGNS were used for both PNPP and GGNS contain-ment response analyses and are shown in Tables 1, 2 and 3 of { Appendix A. The MARCH releases based on Grand Gulf are conser-vative for PNPP, since there was no reduction in hydrogen re-g leases to account for the fact that PNPP has fewer fuel bundles i and less total active cladding to produce hydrogen. f { t _ - - - - - - -
L r L 4.3 CONTAINMENT RESPONSE TO HYDROGEN COMBUSTION l ' CEI, using the hydrogen and steam releases obtained from the MARCH code analyses for Grand Gulf, analyzed the containment L atmosphere transient using the CLASIX-3 coda. The CLASIX-3 code is a modification of the original CLASIX code that was developed to perform hydrogen combustion analyses for an p ice-condenser containment. The CLASIX-3 code is identical to L the CLASIX code in that it is a multivolume containment code, which calculates the containment pressure and temperature re-sponse in the separate compartments. CLASIX-3, however, has [ the capacity to model features of the system unique to a Mark III containment plant (including the suppression pool, refueling pool, vacuum breakers, and drywell purge system). r while tracking the distribution of the atmosphere constituents, ' i.e., oxygen, nitrogen, hydrogen, and steam. The code also has the capacity of modeling containment sprays and structural heat sinks. [ L The CLASIX-3 model for the PNPP analysis is identical to that used for the initial GGNS analyses and sensitivity studies submitted by HCOG letter HGN-001, dated January 15, 1982. [ A diagram of the Mark III containment and a schematic dia-gram of the Perry CLASIX-3 model used in this analysis are { given in Appendix A, Figures 1 and 2, respectively. There are three compartments in this model: the drywell, wetvell and r containment. Also included are the suppression pool, contain-L ment spray system, upper pool, and drywell purge system. The arrows in Figure 2 represent flow paths between compartments with the arrow pointing in the direction of allowed flow. ( Mass and energy released to the containment atmosphere in the form of steam and hydrogen are input to the code. The burning of hydrogen is calculated in the code with provision to [ vary the conditions under which hydrogen is assumed to burn and conditions at which the burn will propagate to other compart-ments. Two CLASIX-3 runs were made for the PNPP. The input for these two cases was identical except for suppression pool drawdown and the location of the steam, hydrogen and fission [ In the stuck open relief valve (SORV) product energy releases. case, the releases entered directly into the wetwell side of the suppression pool over the entire transient. Twenty minutes [ into the transient, the igniters and two Combustible Gas Con-trol System (CGCS) compressors were manually activated and r began pumping gasses from the containment to the drywell. ( After thirty minutes into the transient, the upper pool began dumping water to the suppression pool through one line and con-tinued dumping for 8.67 minutes. The drawdown of the suppres- [ sion pool (reinstatement of injection systems) was initiated at t - - - - - - - -
I 6500 seconds into the transient. Releases in both cases were continued until hydrogen equivalent to a 75% fuel clad metal-water reaction was released from the primary system. At I this time, the SORV transient was terminated. The DWB tran-sient was continued in order to allow the remaining hydrogen to burn although the concentration was less than 8 v/o. Appendix A to this report provides a detailed description of the CLASIX-3 model, input assumptions, plant-specific I parameters used and the results. A summary of the results of the two PNPP cases is given in Table 17 of Appendix A. Temper-ature and pressure information is given in Figures 3-9 for the stuck open relief valve (SORV) case and Figures 22-28 for the I drywell break (DWB) case. Volume fractions of oxygen, nitro-gen, hydrogen, and steam are shown in Figures 10-21 for the SORV case and Figures 29-40 for the DWB case. Table 18 com-I pares the results of two similar analyses performed as part of the sensitivity studies for GGNS. I I !I I lI I I 'I l E
I I 5.0 DESIGN COMPARISON TO GRAND GULF I
5.1 INTRODUCTION
CEI has conducted a significant amount of plant specific analy-ses justifying the PNPP HCS, as described in the previous sec-tions. Further justification is provided by the similarities between PNPP and GGNS hydrogen control systems and containment designs and the fact that the NRC staff has reviewed the GGNS hydrogen control system and approved a full power operating license on an interim basis. This section demonstrates that the GGNS and PNPP designs are similar in all material respects related to hydrogen control. This section provides a design comparison between GGNS and PNPP for: the igniter system, the containment ultimate capacity, the containment systems, the containment response analysis, and the I list of equipment required to survive a hydrogen generation event. This report references FSAR figures and tables for both GGNS and PNPP. A list of FSAR tables cross-referenced to the tables included in this report is provided in Table 5.1-1. 5.2 IGNITER SYSTEM DESIGN The PNPP HCS design is described in detail in section 2.0 of I this report. The GGNS igniter system design has been described in several letters to the NRC; the important aspects of the de-sign are summarized in Supplement 3 to the GGNS Safety Evalua-tion Report (SER), NUREG-0831. A comparison of the most sig-I nificant design features is included in Table 5.2-1. . g PNPP and GGNS have approximately the same number of igniters l g located throughout the drywell, wetwell and upper containment (PNPP has 12 more). The locations of the igniters at the two plants are similar because the same location criteria were used I and because the internal containment configurations are simi-lar, as shown in Figures 1.2-3 through 1,2-10, and Figures 1.2-2 through 1.2-7 of the PNPP and GGNS FSAR's, respectively. The igniter selected for PNPP is a glow plug, Model 7G, manu-factured by General Motors AC Division, and is identical to that installed in GGNS. In both designs the igniter is powered l I directly from a 120/12 VAC transformer. The igniter assembly design is identical to the GGNS assembly and includes the igniter enclosure and the junction box. The igniter enclosure consists of a stainless steel box with 1/8 in. thick walls, which houses the transformer and associated I I
l I electrical connections and partially encloses the igniter. l The sealed box uses a hooded spray shield to reduce water im-pingement on the glow plug. At both GGNS and PNPP, the igniters are powered from Class lE power panels that are supplied from Engineered Safety Features I (ESF) buses through Class lE motor control centers. In the event of a loss of offsite power, the igniters would be powered from the emergency diesel generators. The HCS is designed as a seismic Category I system. At both plants, the HCS is designed so that it can be manually activated from the control room following the start of an acci-I dent, and remain activated until the threat to containment in-tegrity resulting from hydrogen release has passed. PNPP, like GGNS, uses a conservative seven day criterion for duration of I HCS continuous operation. The system has two control switches, one for each electrical division, for actuating the igniters upon indication that the reactor vessel water level has dropped to top of the active fuel. To ensure that the HCS will function as intended, PNPP, like GGNS, is implementing preoperational and surveillance testing I programs. Preoperational testing is performed to verify the proper functioning of controls, wiring, instrumentation, and critical components of the HCS. As at GGNS, testing at PNPP will assure that the surface temperature of the operating ig-niter is equal to or greater than 1700*F. The current in each circuit was measured during preoperational testing at GGNS in order to provide baseline data for determining igniter op-I erability during plant operation. The need for such data depends upon the surveillance requirements in the technical specifications. The PNPP Technical Specifications are cur-I rently under development and will be finalized prior to fuel load. I The actuation criteria for the drywell purge compressors, con-tainment sprays and igniters at GGNS were implemented as pre-liminary procedural requirements prior to development of the generic emergency procedure guidelines. Operations of the HCS I and associated containment systems at PNPP will be in accor-dance with the generic emergency procedure guidelines. 5.3 CONTAINMENT STRUCTURAL CAPACITY Analyses have been performed at both PNPP and GGNS to determine the ultimate structural capacity of their Mark III contain-I ments. The ultimate structural capacity was defined as the pressure at which a general yield state is reached at a criti-cal structural section. I I
CEI determined that the capacity of the PNPP containment shell to resist statically applied pressure is 78.0 psig, based upon the lower bound vessel strength, and 94.0 psig, based upon the mean vessel strength. The most limiting penetration can with-stand 50 psig based on ASME Service Level C limits. Based on more realistic Service Level D limits, the most limiting pene-tration can withstand a 56 psig internal pressure. GGNS determined the ultimate capacity of its Mark III contain-ment by taking into account the strength of the steel liner, and by considering actual steel material strengths. The lower bound vessel capacity was 62 psig and the mean containment ca-pacity was determined to be 67 psig. Based on code specified material strengths, the ultimate capacity was determined to be 56 psig. The most limiting penetration can withstand 56 psig internal pressure. These ultimate containment capacity values for both the PNPP and GGNS, demonstrate a margin of 2 to 3 times the calculated peak containment pressure following a hydrogen burn. The cal-culated peak pressures were 21 psig and 24 psig for PNPP and GGNS respectively, based on the SORV initiating event scenario. 5.4 CONTAINMENT SYSTEMS DESIGN The containment systems relevant to the analysis of the HCS in-clude the containment structure, containment heat removal sys-tems, combustible gas control system, and the suppression pool I make-up system. I 5.4.1 Containment Structure I Both PNPP and GGNS are Mark III pressure suppression contain-ments. The internal arrangement of major equipment and struc-tures are similar as indicated in GGNS FSAR Figures 1.2-2 to 1.2-7, and PNPP FSAR Figures 1.2-3 to 1.2-10. The major difference between the two containment designs is that GGNS has a reinforced concrete containment and PNPP has a I free standing steel containment. Additionally, the PNPP con-tainment is slightly smaller due to the lower reactor power level of 3579 MWt versus 3833 MWt at GGNS. The volumes of the drywell and containment are comparable as indicated below: I I
Drywell (ftd l 3 Containment (ft_y PNPP 277,685 1,141,014 GGNS 270,000 1,400,000 Both containments are designed for 15 psig and 185 F. GGNS is ! designed for 3.0 psid external pressure differential, while PNPP is designed for 0.8 psid external pressure differential. However, PNPP has redundant safety related containment vacuum breakers to maintain pressure within the design external dif-I ferential pressure. Other key containment design features shared by GGNS and PNPP l I are shown in PNPP FSAR Tables 6.2-1 through 6.2-9 and GGNS FSAR Tables 6.2-1 through 6.2-9, which are included as Tables 5.4-1 through 5.4-9 and 5.4-10 through 5.4-18, respectively. 5.4.2 Containment Heat Removal System The containment heat removal system, consisting of the suppres-sion pool cooling and containment spray systems, is an integral l I part of the Residual Heat Removal (RHR) system at both GGNS and PNPP. The purpose of this system is to prevent excessive con-tainment temperatures and pressures to maintain containment in-I I tegrity following an accident. To fulfill this purpose, the containment heat removal systems at both GGNS and PNPP meet the following safety design bases: , a. The system shall limit the long term bulk tem-perature of the suppression pool to 185'F with-out spray operation when considering the energy l I b. additions to containment following a LOCA. The single failure criteria applies to the sys-tem.
- c. The system is designed to safety grade require-ments including the capability to perform its function following a Safe Shutdown Earthquake.
l g d. The system shall maintain operation during those g environmental conditions imposed by the LOCA.
- e. Each active component of the system is testable I during normal operation of the nuclear power plant.
During system operation, water is drawn from the suppression I pool, pumped through one or both RHR heat exchangers and deliv-ered to the suppression pool or to the containment spray I I
I header. Water from the safety service water systems is pumped through the tube side of heat exchangers to cool the suppres-sion pool water. At both plants, the containment spray system can be started manually or automatically. The containment spray system is in-itiated automatically on high containment pressure of 9 psig, with an interlock to delay initiation until 10 minutes after a LOCA initiation signal. The important design parameters for the GGNS and PNPP contain-ment heat removal system are provided in Tables 5.4-11 and 5.4-3, respectively. The key parameters are comparable, dif-fering only to account for the smaller PNPP reactor power level. The piping design of the RHR system for PNPP and GGNS is shown in FSAR Figures 5.4-13 and 5.4-14, respectively. The design is essentially the same with the exception that PNPP in-cludes an additional isolation valve on each subsystem. I Although not part of the containment heat removal systems, the key design features of the Emergency Core Cooling Systems (ECCS) are provided in Tables 5.4-20 for GGNS and 5.4-19 for PNPP. The piping design for the High Pressure Core Spray (HPCS) and Low Pressure Core Spray (LPCS) are shown in FSAR Figures 6.3-1, 6.3-4 for GGNS, and 6.3-7, 6.3-8 for PNPP. The piping design and design values are essentially the same, sized appropriately for the smaller PNPP reactor power. 5.4.3 Combustible Gas Control The combustible gas control system (CGCS) ir provided to con-trol the concentration of hydrogen which may be released in the drywell and containment following a postulated design basis accident (LOCA). At both GGNS and PNPP, the system is composed I of three major subsystems: drywell purge, hydrogen recombiner, and a backup containment purge system. Since the backup con-tainment purge system is not relevant to degraded core hydrogen control, only the first two subsystems will be discussed. The key design and performance characteristics of the GGNS and PNPP CGCS's are provided in Tables 5.4-21 and 5.4-22, respectively. The piping diagrams are shown in FSAR Figures 6.2-81 for CGNS and 6.2-62 for PNPP. For both GGNS and PNPP, the drywell purge subsystem consists of two redundant 100% capacity compressors and associated compo-nents. The compressor draws air from the containment volume and discharges into the drywell, causing flow of the drywell atmosphere through the horizontal vent system, through the sup-pression pool and back into the containment. The only signifi-cant difference between the two designs are: I I
I , I a. PNPP compressors are rated at 546 scfm versus the 500 scfm (minimum) at GGNS (1000 scfm per GGNS Technical Specification 3/4.6.7.3).
- b. The PNPP drywell purge system is manually oper-ated, while the GGNS system is initiated either I manually or automatically (LOCA signal and drywell pressure within 1.0 psid of containment pressure) due to the additional function of post-LOCA drywell vacuum relief.
- c. The PNPP drywell purge discharge penetrates the drywell through the same penetration as the drywell vacuum breakers. The GGNS design has separate penetration for the post-LCCA vacuum breaker lines.
- d. The GGNS drywell purge discharge lines include vaccum breakers for additional vacuum relief once the system is initiated. The PNPP design does not include this feature.
None of the differences identified above would have a signifi-cant effect on the analysis of the HCS. Plant specific differ-ences in system design values were included in the containment analysis as discussed in sections 4.0 and 5.5. Both GGNS and PNPP include two 100%-capacity hydrogen recombiners inside the containment. The hydroger. recombiners are thermal recombiners manufactured by Westinghouse, each having a capacity of 100 scfm and a power rating of 75KW. The hydrogen recombiner subsystem designs for both PNPP and GGNS are similar. ! 5.4.4 Suppression Pool Makeup System The designs of the Suppression Pool Makeup System (SPMS) are , essentially the same at GGNS and PNPP. The SPMS provides water from the upper containment pool to the suppression pool by gravity flow following a design basis accident (LOCA). The piping system consists of two lines, with two normally closed motor operated valves in series in each line. The piping dia-gram for each system is shown in PNPP FSAR Figure 6.2-67 and GGNS FSAR Figure 6.2-82. Both GGNS and PNPP systems are initiated either manually or au-tomatically following LOCA signals and low-low suppression pool l water level or 30 minutes, whichever occurs first. The quanti-ty of water added to the suppression pool is approximately 36,400 and 32,800 cubic feet for GGNS and PNPP, respectively. L I
For both PNPP and GGNS, the SPMS volume is drained down in less than 10 minutes. For both plant designs, the SPMS will accom-plish its safety function prior to the generation of signifi-I cant amounts of hydrogen. Therefore, the minor system differ-ences discussed above are not pertinent to this evaluation. 5.5 CONTAINMENT RESPONSE ANALYSIS GGNS and PNPP used the CLASIX-3 computer code to evaluate the containment pressure and temperature response to hydrogen deflagration. Both plant analyses used the hydrogen and steam releases obtained from the MARCH code. The MARCH release rates used for PNPP, were conservatively overstated by using the GGNS hydrogen and steam release rates without any reduction to ac-I count for the smaller core size (PNPP has 748 fuel bundles ver-sus the 800 at GGNS). Both GGNS and PNPP evaluated the results of two types of hydro-gen generation events: the more probable transient initiated stuck open relief valve (SORV) event and a drywell small line break case (DWB). The PNPP analysis is included as Appendix A. I The comparable GGNS analysis (cases sal and DA4) was submitted to the NRC as part of the CLASIX-3 sensitivity studies, by HCOG Both plant analyses letter HGN-001, dated-January 15, 1982. use essentially the same model and input assumptions, adjusted for plant specific containment parameters. In addition to using the exact s&me MARCH hydrogen and steam release histories, the key input assumptions used by both analyses in-l cluded: a) Igniters and drywell purge system activated at 20 I minutes into the transient. b) Only one of two containment spray trains initiated after the first hydrogen burn. t c) Burn parameters of:
- 1) H2 V/F for ignition 0.08
- 2) Hx V/F for propogation 0.08
- 3) H2 fraction burned 0.85
- 4) Minimum 2{/Fforignition Minimum 0 V/F to support 0.05 l 5) i combustion 0.00
- 6) Flame speed 6 ft./sec.
d) 50/50 split for LOCA vent /SRV in DWB cases. i e) Suppression pool drawdown. I i
s I f) Initiation of drywell spray (simulating water from the small line break as a coarse spray) with an ini-tial temperature of 175 F. g) Hydrogen release equivalent to 75% metal water reac-tion of active fuel cladding at GGNS. A comparison of the results of the PNPP and GGNS cases is pro-vided as Table 18 of Appendix A. Figures 3 through 40 of Ap-pendix A provide the plotted results of the PNPP CLASIX-3 anal-ysis. The comparable results for GGNS are included as Figures 5.5-1 through 5.5-57. The results of both the PNPP and GGNS SORV analyses are simi-lar. The containment volume at PNPP is smaller than at GGNS by 23%. This contributes to the extra containment burn in the PNPP transient. However, PNPP has a 20% larger initial wetwell volume than GGNS, which results in fewer wetwell burns in the PNPP transient. For the PNPP SORV case, peak temperatures and pressures oc-curred in all compartments during the first of the two contain-ment burns, at approximately 6900 seconds into the transient. The first containment burn resulted in the most severe pressure and temperature excursion because wetwell ignition occurred just before and during the containment burn. No wetwell igni-I tion occurred during the second containment burn due to a lack of oxygen. Four additional wetwell temperature peaks (at approximately 4445, 6555, 6965, and 7220 seconds) are notable. Sprays are not initiated until after the first wetwell burn, which explains why the first wetwell temperature peak is higher than I those which immediately follow. The other three above average wetwell temperature peaks occur because ignition takes place at increased hydrogen concentrations due to insufficient oxygen I concentration when the hydrogen concentration reached the 87 /o setpoint. Peak pressures and temperatures for the PNPP SORV case are com-parable in magnitude to those of the GGNS SORV case, except for the wetwell peak temperature. The PNPP wetwell temperature is higher due to the coincident combustion in the wetwell and con-I tainment, which did not occur in the GGNS SORV case. The results of the PNPP DWB case are also similar to the corre-I sponding GGNS case. Again, fewer vetvell burns are evident for the PNPP.DWB case due to the larger wetwell volume. The only other notable difference between the results for the two plants I relates to the containment burn. The PNPP DWB case originally did not have a containmentThe dryvell and wetvell burn. burn associated with the final volume fraction of hydrogen in I the containment just prior to the final burn was 0.065. The volume fraction of hydrogen required for ignition is 0.08. To be conservative, it was decided to force a containment burn at I this point to obtain peak temperatures and pressures. This re-duced concentration forced burn resulted in lower peak tempera-tures and pressures for the PNPP DWB case. The total amount of hydrogen burned in the PNPP SORV transient was 2011 lbs. and in the DWB transient was 2290 lbs. These values correspond to 77.0% and 87.6%, respectively, of the I total amount of hydrogen that was available. The similar GGNS cases show the SORV case burning 2332 lbs. and the DWB case burning 2243 lbs., which are 89.3% and 85.8% of the total hy-drogen releases, respectively. The difference between the per-centage of hydrogen burned in the PNPP and GGNS SORV cases is due to the greater number of wetwell burns in the GGNS case. I In summary, the CLASIX-3 model, MARCH hydrogen and steam re-lease rates, and key input assumptions were essentially the I same for the PNPP containment analysis and GGNS cases sal and DA4. A comparison between the PNPP and GGNS analysis shows the SORV and DWB transients to be substantially similar. The only notable differences are in peak temperatures and pressure, which are explained by plant geometry, the forced containment burn at a lower hydrogen concentration in the PNPP DWB case, and coincident combustion in the wetwell and containment during I the PNPP SORV case. Other than these differences discussed above, the burn temperature and pressures are approximately the same. In addition for PNPP, the number of burns in the wetwell I is less than GGNS, with more spacing between burns. For both PNPP and GGNS the peak pressures are well within structural ca-pability. For PNPP, the increase in the peak burn temperature over GGNS will have little effect on equipment survivability I due to the short duration of each burn. Further, fewer burns in the wetvell for PNPP with approximately the same peak tem-perature as GGNS should result in a lower average temperature and a lower equipment temperature. 5.6 SURVIVABILITY OF ESSENTIAL EQUIPMENT i A consequence of controlling excessive hydrogen generation by deliberate ignition is high peak containment atmosphere temper-atures. A preliminary evaluation has been performed to identi-fy equipment inside containment required to survive a hydrogen I burn. The identification of the equipment that has to survive the hy-drogen burn environment was based on its function during and after postulated degraded core accidents. In general, equip-I ment located in the containment in the following four I I
a categories was considered to be essential for safety of the plant:
- 1. systems mitigating the consequences of the acci-dent;
- 2. systems needed for maintaining integrity of the containment pressure boundary; systems needed for maintaining the core in a I 3.
safe condition;
- 4. systems needed for monitoring the course of the accident Using these criteria, for the preliminary evaluation PNPP has I prepared a list of equipment inside containment and dryvell re-quired to survive a hydrogen burn. This list is presented in Table 5.6-1 and 5.6-2 for PNPP drywell and containment equip-I .ae n t , respectively. For comparative purposes, the GGNS equip-ment survivability lists which were provided to NRC by letter dated October 17, 1983 (AECM-83/0671) are included in Tables 5.6-3 and 5.6-4.
Both plant lists contain similar components that have similar functions. Evaluation of the equipment required to survive hy-I drogen combustion events against- the thermal environment will be addressed in the final analysis. I lI
- I I
lI l I l
Table 5.1-1 Preliminary Evaluation PNPP/GGNS Table No. FSAR Reference I 5.2-1 N/A 5.4-1 PNPP Table 6.2-1 5.4-2 PNPP Table 6.2-2 I 5.4-3 PNPP Table 6.2-3 5.4-4 PNPP Table 6.2-4 5.4-5 PNPP Table 6.2-5 5.4-6 PNPP Table 6.2-6 5.4-7 PNPP Table 6.2-7 I 5.4-8 PNPP Table 6.2-8 5.4-9 PNPP Table 6.2-9 5.4-10 GGNS Table 6.2-1 5.4-11 GGNS Table 6.2-2 5.4-12 GGNS Table 6.2-3 5.4-13 GGNS Table 6.2-4 l 5.4-14 GGNS Table 6.2-4 l 5.4-15 GGNS Table 6.2-5 5.4-16 GGNS Table 6.2-6 5.4-17 GGNS Table 6.2-7 5.4-18 GGNS Table 6.2-8 5.4-19 PNPP Table 6.3-1 5.4-20 GGNS Table 6.3-2 5.4-21 GGNS Table 6.2-45 I lI L
I 5.4-22 PNPP Table 6.2-37 5.6-1 & 5.6-2 N/A (PNPP Equipment Lists) 5.6-3 & 5.6-4 N/A (GGNS Equipment Lists) I I I I I I lI I I I I I l I I
TABLE 5.2-1 PNPP/GGNS HCS Design Comparison I 1. Number of Igniters o Drywell PNPP 17 GGNS 18 12 11 I o o o Wetwell Enclosed Area Containment 22 51 16 45 Total 102 90 I 2. Igniter Location Criteria (except drywell below weir wall and containment above 1 ESF Division operable: 60 ft not to exceed 70 ft. 60 ft. not to exceed 70 ft. I refueling floor) 2 ESF Divisions operable: 30 ft. not to 30 ft. not to exceed 35 ft. exceed 35 ft.
- 3. Igniter Assembly Power Systems Power Systems I Maufacturer Division of Morris Knudson Division of Morris Knudson GMAC Model 7G GMAC Model 7G I
- 4. Igniter Operating 1700* 12 VAC 1700*F 12 VAC Temperature
- 5. Igniter Transformer 0.2 KVA Dongan 0.2 KVA Dongan Model 52-20-472 Model 52-20-472
; 6. Igniter Qualification In progress 330*F for 3 hrs.
Temperature
- 7. Igniter Qualification In progress 70 psig Pressure
- 8. System Operation Manually via 2 Manually via 2 l
I control room handswitches (1 switch per control room handswitches (1 switch per lI
I division) division)
- 9. Power Supply 120 VAC 1 10% 120 VAC 1 10%
from ESF Power from ESF Power Panels powered Panels powered off of motor off of motor control Centers control centers from ESF buses from ESF buses (on site and I (on site and offset AC power supplies) offset AC power supplies) I I I ^ I I - I I I I I I
- I I
I TABLE 5.4-1 I KEY DESIGN AND MAXIMUM ACCIDENT PARAMETERS FOR PRESSURE SUPPRESSION CONTAINMENT I Parameter Design Value Maximum Calculated Accident Value Containment Pressure, psig 15 12.0 Containment Temperature, *F 185 184.6 I Drywell Pressure, psig 30 22.1 Drywell Temperature, *F 330 330 I I I I I I I I I I I I
I TABLE 5.4-2 l I ' CONTAINMENT DESIGN PARAMETERS Devuell containment Drywell and Containment Negative Design Pressure, psig -21.0 -0.8 Positive Design Pressure, psig 30 15 Design Temperature, OF 330 185 Net Free Volume, ft3 277,685 1,141,014 Maximum Allowable Leak Rate 5,843 SCFM 0 2.5 psig 0.2%/ day 32,645 SCFM g 30 psig Suppression Pool Water Volume, fc3 Low Level 11,155 105,950 High Level 11,395 108,750 Suppression Pool Surface Area, ft2 , 482 5,900 Suppression Pool Depth, ft Low Level 18.0 18.0 High Level 18.5 18.5 Upper Pool Makeup Volume, ft3 - 32,830 I I I I I I >
l I TABLE 5.4-2 (continued) I I Containment Vent System 120 Number of Vents Nominal Vent Diameter, ft 2.29 Total Vent Area, ft 495 I Vent Centerline Submergence (low level), ft Top Row 7 Middle Row 11.5 Bottom Row 16 Vent Loss Coefficient (varies with number of vents open) 2.5-3.5 I I I I I I
; I I
I I
I TABLE 5.4-3 ENGINEERED SAFETY FEATURE SYSTEMS PERFORMANCE PARAMETERS FOR CONTAINMENT RESPONSE ANALYSES I Containment Spray Full Capacity Containment Analysis Value Case A Case B Number of RHR Pumps 2 0 0 Number of Lines 2 0 0 Number of Heaters 2 0 0 Flow Rate, gpm/ pump 5250 0 0 Containment Cooling System Number of RHR Pumps 2 2 1 Pump Capacity, gpm/ pump 7100 7100. 7100 RHR Heat Exchangers Type Inverted U-tube, single pass shell, multipress tube, vertical mounting Number 2 2 1 Heat Transfer Area,f t / unit 14,850 - - Overall Heat Transfer Coefficiegt, Btu /hr-ft *F/ unit 200 - - Service Water Flow Rate, gpm/ unit 7300 7300 7300 Service Water Temperature,*F 5 Minimum Design 32 - - Haximum Design 80 80 80 I Containment Heat Removal Capability (using 80*F service water and 185* pool tenperature) 6
- Btu /hr/ unit 166.4x10 - -
lI I
1 l TABLE 5.4-4 ACCIDENT ASSUMPTIONS AND INITIAL CONDITIONS FOR CONTAINMENT RESPONSE ANALYSES I Components of Effective Break Area (recirculation line break), ft' Recirculation Line 2.127 Cleanup Line 0.062 Jet Pumps 0.461 Primary Steam Energy Distribution (I) , 100 Btu Steam Energy 25.59 Liquid Energy 722.3 Sensible Energy Reactor Vessel 98.25 Reactor Internals (less core) 40.49 Primary System Piping 45.40 Fuel (2) 7.2 I Other Assumptions Used in Analysis Main Steam Closure Time, sec Recirculation Break 3.5 I Main Steam Line Break 5.5 Scram Time, sec <1 I NOTES: I 1. All energy values, except fuel, are based upon a 32*F datum.
- 2. Fuel energy is based upon a datum of 285*F.
I I
TABLE 5.4-5 I INITIAL CONDITIONS EMPLOYED IN CONTAINMENT RESPONSE ANALYSES Reactor Coolant System (1) Reactor Power Level, MWt 3,651 Average Coolant Pressure, psia 1,040 Average Coolant Temperature, *F 549 Mass of Reactor Coolant System Liquid, Ibm 544,540 Mass of Reactor Coolant System Steam, Ibm 21,530 Volume of Liquid in Reactor Pressure Vessel, ft 11,838.3 3 Volume of Steam in Reactor Pressure Vessel, ft 9,189.2 3 Volume of Liquid in Recirculation Loops, ft 742 3 1,454 Volume of Steam in Steam Lines, ft Volume of Liquid in Feedwater System, ft 3 24,303 3 Volume of Liquid in Miscellaneous Lines, ft 84 I Drywell and Containment Drywell Containment Pressure, psig 0 0 Air Temperature, *F 135 90 Relative Humidity, % 40 50 Suppression Pool Water Temperature, *F 90 90 3 Suppression Pool Water Volume, ft 8,680 105,950 Top Row Vent Centerline, ft 7.0 7.0 I I I
l TABLE 5.4-5 (continued) Dryvell and Containment (Cont'd) Drywell Containment Upper Pool Water Temperature, 'F - 100 3 Upper Pool Makeup Water Volume, ft - 32,830 l l NOTE: i I 1. Reactor coolant system at 102 percent of rated power and normal liquid levels. I I I I I I I I I I
TABLE 5.4-6 ! l I
SUMMARY
OF SHORT TERM CONTAINMENT RESPONSES TO I RECIRCULATION LINE AND MAIN STEAM LINE BREAKS (MINIMUM ECCS) Recirculation Main Steam Line Break Line Break Peak Drywell Pressure, psig 21.26 22.1 Peak Drywell Differential Pressure, psid 20.26 21.05 Time of Peak Pressure, see 1.89 1.8 Peak Drywell Temperature, 'F 248.8 324 Peak Wetwell Pressure, psig 9.82 10.36 Time of Peak Wetwell Pressure, see 462.5 691.6 Peak Suppression Pool Temperature during Blowdown, 'F 155.8 157.8 Calculated Drywell Margin, % 29 26.33 Energy Released to Contaigment at Time of Peak Pressure, 10 Btu 9.0 9.0 Energy Absorbed by Passive Hgat Sinks at Time of Peak Pressure, 10 Btu 0 0 I 1 I lI I
- I lI
TABLE 5.4-7
~
I
SUMMARY
OF LONG TERM CONTAINMENT RESPONSES TO RECIRCULATION LINE OR MAIN STEAM LINE BREAK Case A Case B Peak Containment Pressure, psig 8.58 11.31 Time of Peak Containment Pressure, sec 4,167 11,128 Peak Suppression Pool Temperature, 'F 170.5 184.6 Calculated Containment Margin, % 42.8 24.6 HPCS Flow Rate, gpm 6,000 6,000 LPCS Flow Rate, gpm 7,100 7,100 RHRS Flow Rate, gpm 14,200 7,100 I l l l I I I I I I I I
I TABLE 5.4-8 I ENERGY BALANCE FOR MAIN STEAM LINE BREAK ACCIDENT Energy (BTU) Initial Drywell End of Long Term Peak (time zero) Peak Pressure Blowdown Wetwell Pressure Reactor Coolant 3.2+8 3.1+8 7.5+7 2.0+8 Fuel and Cladding Fuel 7.2+6 7.2+6 0 0 Cladding 3.4+6 3.4+6 1.7+6 1.2+6 Core Internals 1.0+8 1.0+8 9.9+7 3.6+7 Reactor vessel Metal 9.1+7 9.1+7 8.8+7 3.1+7 Reactor Coolant System Included in " Core Internals", above. Piping, Pumps, and Valves Blowdown Enthalpy 9.1+5 7.8+8 4.4+9 I Liquid Steam 0 0 1.0+7 9.9+7 9.9+7 s m E Decay Heat 0 3.0+6 8.6+7 7.5+8 Metal-Water Reaction Heat 0 1.4+4 1.6+6 1.6+6 Drywell Structures 0 0 0 0 Drywell Air 1.8+6 2.1+6 1.2+0 1.3+6 Drywell Steam 8.4+5 1.0+7 2.1+7 9.0+6 Containment Air 7.6+6 7.7+6 1.0+7 9.5+6 Containment steam 1.3+6 2.5+6 1.5+7 2.6+7 Suppression Pool Water 4.2+8 4.2+8 1.1+9 1.2+9 Upper Pool Dump 1.4+8 1.4+8 1.4+8 0 I Inventory 3.9+8 i Energy Transferred 0 i 0 0 by Heat Exchangers Passive Heat Sinks 0 0 0 0 I I .
TABLE 5.4-9 s' ACCIDENT CHRONOLOGY FOR 1 MAIN STEAM LINE BREAK ACCIDENT l Time (sec) I Event All ECCS in Operation Minimum ECCS Available First Row Vent Cleared 0.897 0.897 Second Row Vent Cleared 1.104 1.104 Third Row Vent Cleared 1.511 1.511 Drywell Reaches Peak Pressure 1.8 1.8 Maximum Positive Differential Pressure Occurs 1.8 1.8 Initiation of ECCS Operation ,30 30 Third Row Vent Recovered 32 32 Second Row Vent Recovered 53 53 End of Blowdown 322 416 Reactor Pressure Vessel Reflooded 309 696 First Row Vent Recovered 730 730 I Initiation of RHR Heat Exchanger Operation 1,980 1,980 Containment Peak Pressure Reached 4,167 11,128 I I I I I
I 1 I
- TABLE 5.4-10 CONTAINMENT DESIGN PARAMETERS Drywell Containment A. Drywell and Containment Internal pressure, psig 30 15 External design pressure differential, psid 21 3.0 Design temperature, F 330 185 I Net free volume,, ft 270,000 1,400,000 I Maximum allowable leak rate, %/ day NA .35%**
Suppression pool water volume I Minimum, ft 13041* 122250 Maximum, ft 13303* 125398 2 6666 Pool cross-section area, ft 554 18'7" Pool depth'(normal) 18'7" I Including horizontal vents Combining Based on containment free air volume @ 11.5 psig. this value with the MSIV leakage criteria of 100 scfh (total for four steam lines) yields an everall leakage criteria, based on total volume of containment and drywell, of 0.437%/ day. I
)
I .
TABLE 5.4-10 (continued) t Containment B. Vent System l
- 1. No. of vents 135
- 2. Nominal vent diameter, ft 2.33 ,
- 3. Total vent area, ft (gross) 577.3
- 4. Net vent area, ft (unobstructed) 552.0
- 5. Vent centerline elevation Top row 11'4" Middle row 7'2" Bottom row 3'0" Pool bottom (assumed datum) 0'0"
- 6. Vent loss coefficient (fL/D)
Varies with the number of vents open 2.5 - 3.5 I
i TABLE 5.4-11 ENGINEERED SAFETY SYSTEMS INFORMATION FOR CONTAINMENT RESPONSE ANALYSES - Full Containment Analysis Value Capacity Case A Case B A. Suppression Pool Cooling (RHR system)
- l. No. of pumps 2 2 1
- 2. No. of lines 2 2 1 l.
- 3. Flow rate, gpm/ pump 7450 7450 7450 i
B. Emergency Coolinct Water System
^
- 1. Number of pumps 2 2 1
') 2. Flow capacity, gpm/ loop min 7450 7450 7450
- 3. RHR heat exchangers l
) a. Type - Inverted l U-tube, single !g pass shell, multi-lg pass tube, vertical mounting
- b. Number 2 2 1 l
'* Cases A and B defined in Table 6.2-6 ~
\
i I TABLE 5.4-11 (continued) I Full 9entainment Analysis Value Capacity Case A Case B B.3 (Cont. )
- c. Heat transfer 21250 area, ft 2/ unit 21250 21250
- d. Overall heat transfer coefficient Btu /hr - ft* -
F 212 I e. Secondary coolant flow rate per 3.95 x 10 6 3.95 x 10 6 3.95 x 10 6 exchanger, lb/hr
- f. Design standby service water temperature Maximum, F 90 90 90 Minimum, F 40 I g. Containment heat removal capability per loop, using 90 F
- service water and I 185 F pool tempera-ture; and at rated 6
184.7 x 10 Btu /Hr C. ECCS System
- 1. High pressure core spray (HPCS)
No. of pumps 1 1 1 a. 1 1 1
- b. No. of lines 7115 7115 7115
- c. Flow rate, gpm Low pressure core I
2. spray (LPCS) 1 1 1
- a. No. of pumps 1 1 1
- b. No. of lines Flow rate, (rated, gpm/line) 7115 7115 7115 c.
- 3. Low pressure coolant injection (LPCI)
No. of pumps 3 3 1 a. I Sheet 2 of 3
I TABLE 5.4-11 (continued) Full Containment Analysis Valt I Capacity Case A Case B l ' l E C.3 (Cont.) 3 3 1
- b. No. of lines Flow rate, gpm/line 7450 7450 7450 c.
D. Automatic Depressurization System l
- 1. Total number of safety /
relief valves 20 8
- 2. No. actuated on ADS lI I
I I ( l
- I 1
I 1 I i Sheet 3 of 3 E
TABLE 5.4-12 I ACCIDENT ASSUMPTIONS AND INITIAL l COND1TIONS FOR LARGE LINE BREAKS I A. Effective accident break area (total), recirculation line break, ft2 3.181 I B. Effective accident break area, main steam line break, ft2 3.538 Components of effective break area (recirculation I C. line break): 2.598
- 1. Recirculation line area, ft I 2. Cleanup line area, ft
.080 503
- 3. Jet pump area, ft
( } D. Primary steam energy distribution I 1. Steam energy, 10 6 6 Btu 31.4 341.4
- 2. Liquid energy, 10 Btu
- 3. Sensible energy, 10 Btu 111.0
- a. Reactor vessel 58.1
- b. Reactor internals (less core) 37.7
- c. Primary system piping 27.6
- d. Fuel (2)
- c. Other assumptions used in analysis I 1. Deleted 5.5
- 2. MSIV closure time (sec)
<1
- 3. Scram time (sec) 100
- 4. Liquid carryover,,%
I III All energy values except fuel are based on a 32 F dacum. I> Fuel energy is based on a datum of 235 F. I
TABLE 5.4-13 INITIAL CONDITIONS EMPLOYED , IN CONTAINMENT RESPONSE ANALYSES I A. Reactor Coolant System (at design overpower of 105% and at normal liquid levels)
- 1. Reactor power level, MWT 3995 Average coolant pressure, psia 1060 2.
- 3. Average coolant temperature, F 551 5
- 4. Mass of reactor coolant system liquid, lbm 6.815 x 10 Mass of reactor coolant system steam, lbm 24,000 5.
I 6. Liquid plus steam energy, Btu 372.8 x 10 6 Volume of liquid in vessel, ft 3,771 7. Volume of steam in vessel, ft 9,295 8. Volume of liquid in recirculation loops, ft 827 9. 3 25,820
- 10. Total reactor coolant volume, ft I
lI l - lI l I I I l Sheet 1 of 2
TABLE 5.4-14 I'B. Containment Drywell Containment
- 1. Pressure, psig 0.0 0.0 t
3 ! 3 2. Inside temperature 135 95
- 3. Relative humidity, % 20 to 90 60 4 Service water 90 90 temperature, F I
I I I - 1g 1 I I I I , I I-Sheet 2 of 2 I
I ' TABLE 5.4-15 I
SUMMARY
OF SHORT-TERM ACCIDENT RESULTS FOR CONTAINMENT RESPONSE TO RECIRCULATION LINE AND STEAM LINE BREAKS Accident Parameters I A. Recirculation Steam Line Break Line Break I 1. Peak drywell pressure, psig 19.4 22.0 J
- 2. Time (s) of peak pressures, sec 1.09 1.09 l53
- 3. Peak drywell temperature ,
F 240 330
- 4. Peak suppression pool temperature during blow-dcwn, F , 120 120
- 5. Calculated drywell margin, % 35 27
- 6. Energy released to con-tainment at time of short- 53 term peak pressure, 106 Etu 240 240
- 7. Energy absorbed by passive heat sinks at time of peak pressure, 106 Btu 0 0 I II See Figures 6.2-2 and 6.2-5 for plots of pressures vs time.
See Figures 6.2-3 and 6.2-7 for plots of temperatures vs time. I I' I
TABLE 5.4-16 T.OSS OF COOLANT ACCIDENT LONG TERM Pit 1 malty CONTAINMENT llisSPONSE
SUMMARY
Service Containment Secondary LPCI/LPCS Water Spray IIPCS LPCI/LPCS Peak Peak Pressure Case
- Pumps _ Pumps (gal / min) (gal / min) (gal / min) Pool Temp. F (psig) 3/1 2 0 7115 7450/7115 155.5 7.6 A
B 1/1 1 0 7115 7450/7115 171.3 9.9 l l
*A - Assumes offsite power available B - Assumes loss of offsite power E -
M M M M M M M M M m m a e e
TABLE 5.4-17 ENERGY BALANCE FOR DESIGN BASIS RECIRCULATION LINE BREAK Energy Levels vs Time (Minimum ECCS - Misgile Break) Energy in 10 Btu Maximum Initial Peak op End Blowdown Containment Pressure Parameter (t = 0) (t=1.1865 sec) (t=130.62 sec) (t=18305.7 sec) Reactor coolant 390.0 377.0 29.2 174.0 Fuel 39.6 40.3 5.86 3.75 Cladding 3.14 3.14 1.30 0.831 Iteactor vessel _101.0 101.0 88.3 26.7 Reactor internals 96.3 96.3 85.9 25.5 Drywell air 1.70 2.02 ~0. 1.31 Drywell steam 0.817 13.3 16.8 3.76 Drywell liquid 0, 1.33 26.7 530.0 Containment air 8.95 9.14 11.3 8.53 Containment steam 3.61 3.63 9.98 24.1 Containment liquid suppression 1200. 1200. 1610. 1370. pool Decay heat O. 0.380 23.5 932.0 Metal water heat O. ~0. 0.035 0.463 Pump heat 0. O. 0.512 88.3 Ileat transferred O. O. O. 634 RilR heat exchanger M M M M M M M M M M M M
I TABLE 5.4-18 i ACCIDEUT CERONOLOGY-DESIGN BASIS RECIRCULATION LINE EREAR ACCIDE!iT Time (sec) Case A Case B All ECCS Min ECCS Event in Operation Available
- 1. 1st row vent cleared .86 .86
- 2. 2nd row vent cleared 1.08 1.08
- 3. 3rd row vent cleared 1 44 1 44
- 4. Drywell reaches peak pressure 1.09 1.09 I 5. Maximum positive differential pressure occurs 1.08 29 29 1.08
- 6. 3rd row vent recovered
- 7. Initiation of the ECCS 30 30
- 8. 2nd row vent recovered 40 40
- 9. 1st row vent recovered 99 99 99 99
- 10. End of blowdown
- 11. Vessel reflooded 279 455
- 12. Initiation of RER heat 1800 -
1800 exchanger loop
- 13. Containment reaches peak pressure 4936 23176 I
I I I I' I
TABLE 5.4-19 SIGNIFICANT INPUT VARIABLES USED IN THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Variable Units Value A. PLANT PARAETERS Core thermal power MW e 3729 Vessel steam output lba /hr 16.2x106 I Corresponding percent : 105 of rated steam flow I Vessel steam dome pressure psia 1060 Maximum recirculation line break area ft2 2.7 i B. EMERGENCY CORE COOLING SYSTEM PARAMETERS B.1 Low Pressure Coolant Injection System Vessel pressure at psid (vessel which flow may commence to drywell) 225 I Minimum rated flow at vessel pressure GPM psid (vessel 19500 20 to dryvell) Initiating Signals low water level ft. above >1.0 or top of active fuel high drywell pressure psig 12.0 Maximum allowable time sec 27 delay from initiating signal to pumps at rated speed l Injection valve fully sec. after DBA <40 open I I I
I TABLE 5. 4-19 (continued) I Variable Units Value B.2 Low Pressure Core Sorav Svstem Vessel pressure at which psid (vessel flow may commence to drywell) 289 Minimum rated flow gpm 6000 at vessel pressure psid (vessel to drywell) 122 I Initiatine Sienals low water level ft. above top >1.0 or of active fuel high drywell pressure psig <2.0 Maximum allowed (runout) gpm 7800 flow
- Maximum allowed delay sec 27.0 g ' time from initiating 3 signal to pump at rated speed Injection valve fully sec. after DBA <40 open B.3 High Pressure Core Spray Vessel pressure at which psid 1177 flow'may commence Minimum rated flow avail- gpm 517 1550 5000 I able at vessel pressure psid (vessel to pump suction) 1177 1147 200 Initiating Signals low water level ft. above top >10.9 or of active fuel high drywell pressure psig <2.0 I Maximum allowed (runout) flow gpm 7800 l
I Maximum allowed delay time from initiating signal to rated flov sec 27.0 l 3 available and injection 5 valve wide open t I
TABLE 5.4-19 (continued) Vr.riable Units Value B.4 Automatic Depressurization Svstem Total number of relief valves with ADS function 8 Total minimum flow capacity lb/hr 6.4x106 at vessel psig 1125 pressure Initiating Signals low water level ft. above top >
- 1. 0 and of active fuel high drywell pressure psig < 2.0 Delay time from all see 1 120 initiating signals completed to the time valves are open C. FUEL PARAMETERS Fuel type Initial core Fuel bundle geometry 8x8 Lattice C Number of fueled rods per bundle 62 Peak technical specification kW/ft 13.4 linear heat generation race Initial minimum critical 1.17 j power ratio Design axial 1.4 peaking factor l g
! g I lI (
1 I I GG FSAR TABLE 5.4-20 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters i o Core Thermal Power MWt 3993 1 I o Vessel Steam Output LBm/hr 17.3 x 106 o Corresponding percent of rated percent 105 steam flow I o Vessel Steam Dome Pressure psia 1060 o Maximum Recirculation Line ft2 3.1 Break Area Emergency Core Cooling System Parameters Low-Pressure Coolant Injection System o Vessel Pressure at which flow psid (vessel 225 may commence to drywell) I o Minimum Rated Flow at Vessel Pressure GPM psid (vessel to drywell) 22000 20 o Initiating signals low-low-low water level ft above top 21.0 or of active fuel high drywell pressure psig 52.0 o Maximum allowable time delay sec 27.0 l I from initiating signal to pumps at rated speed i o Injection valve fully open sec after DBA 540.0 Low-Pressure Core Spray System o Vessel pressure at which flow psid (vessel 289 may commence to drywell) 7000 I o Minimum rated flow at Vessel GPM Pressure paid (vessel 122 to drywell i
GG FSAR TABLE 5.4-20 (continued) I o Initiating signals low-low-low water level or ft. above top of active fuel 21.0 high drywell pressure psig 52.0 o Maximum allowed (runout) flow GPM 9100 o Maximum allowed delay time sec 27.0 from initiating signal to pump at rated speed o Injection valve fully open sec after DBA 540.0 High-Pressure Core Spray o Vessel pressure at which flow psid 1177 may commence o Minimum flow available at See vessel to pump suction head Figure 6.3-3 o Initiating signals
! low-low water level ft. above top 210.5 or of active fuel high drywell pressure psig 52.0 o Maximum allowed (runout) flow GPM 9100 o Maximum allowed delay time sec 27.0 from initiating signal to rated flow available and injection valve wide open Automatic Depressurization System o Total number of valves installed 8 o Number of valves used in 8(1) analysis o Minimun Flow Capacity of lb/hr 6.4 x 106 8 valves at vessel psid (vessel 1125 pressure suppression pool)
(1) Additional LOCA analyses in Section 6.3.3.7*.8 with seven ADS i valves justify one ADS valve out of service for an extended period of time. L I
Il GG FSAR TABLE 5.4-20 (continued) , I o Initiating' signals low-low-low water level ft above top 21.0 and of active fuel high drywell pressure psig 12.0 and W o Delay time from all initiating sec 5120 signals completed to the time valves are open I FUEL PARAMETERS I o Fuel typh Initial Core o Fuel Bundle Geometry -- P8x8R I o Lattice C o Number of fueled rods 62 o Peak Technical Specification kw/ft 13.4 Linear Heat Generation Rate o Initial Minimum Critical Power -- 1.17 Ratio l l l
k co' I FSAR TABLE 5.4-21 COMBUSTIBLE GAS CONTROL SYSTEM COMPONENT DESCRIPTION l Drywell Purge Compressors Quantity 2-100% capacity units capacity (minimum), scfm 500 Static pressure, psig 10 I Drive Motor, hp Manufacturer Direct 100 Turbonetics Hydrogen Recombiners Type Thermal Quantity 2-100% capacity units capacity, scfm - air 100 Process rate, scfm - hydrogen 4 (approx) Power required, kW 75 each I Manufacturer Westinghouse Containment Purge Compressor
~
Type Liquid ring Quantity 1 65 I Capacity, scfm Static pressure, psig Drive 10 Direct Motor, hp 15 Manufacturer Nash I I I
I TABLE 5.4-22 COMBUSTIBLE GAS CONTROL SYSTEM EQUIPMENT DESIGN AND PERF01U1ANCE DATA
- a. Combustible Gas Purging Units (Mixing)
- 1. Compressor Centrifugal Max inlet pressure, psia 23.3 Max discharge pressure, psia 29.13 Max inlet temperature, *F 185 Max discharge temperature, *F 238 Relative humidity (inlet), % 100 Capacity, scfm 546 Power requirement, BHP 41
- 2. Heat Exchanger Design Pressure (tube side), psig 500 Air Temperature in/out, *F 238/190 Cooling Water Temp. in/out, *F 140/170
- 3. Material ,
I Compressor Casing cast steel Shroud aluminum J ' impeller 174 PH S.S. Heat Exchanger Tube 304 S.S. I
I TABLE 5.4-22 (continued)
- 4. Manufacturer .Turbonetics
- b. Isolation Valves Type globe Body Bronze Stem Bronze Disc Hardened 304 S.S.
Disc Type Swivel plug Seats Renewable hardened 304 S.S.
- c. Hydrogen Recombiner
- 1. Material Outer Structure Type 300 series S.S.
Inner Structure Inconel 600 Heater Element Sheath Inconel 800 Base Skid Carbon Steel, painted
- 2. Power Maximum, kW 75 Nominal, kW 50
- 3. Capacity, scfm 100 to 120 at 1 atm
- 4. Temperatures Gas in, *F 150 Outlet of heater section, *F 1,150 to 1,400 Exhaust, *F ,
50 above ambient I I
i !I I 1 I /, TABLE 5.4-22 (continued)
- 5. Heaters 4
5 banks Number 2 5.8 2 Max. heat flux, watts /in 1 Max. sheath temperature, *F 1,550 Westinghouse I 6. Manufacturer
- d. Piping
' Carbon Steel Material l
- I
!I lI 4 i !I } i i I I I
I F4 ue s.s-1 g 200.0. . . ..
- ~
l 180.0 l Ie i2
< .i '.
al: : 160.0 : M z : - IN l E ' 14 0.0 : l E - l !
; o,n., .
l 0.0 2.0 4.0 6.0 8.0 TIME (SECONOS) X 1000 GGNS BASE CASE.SORV g ORYWELL TEMPERATURE I I ' Figure 41 I
I Figure 5.5-2 I I 1100.0, ( : I 900.0
.h. ll( i
! C 700.0 I l __ _
I 500.0 I r
* ' ~
^ ^ :rX_
N :
-- -_ 3 I 300.0
= -- 9 l
! Er I i l - '
O.0 2.0 4.0 6.0 8.0 TIME (SECONDS) X 1000 I GCNS BASE CASE SORV I l . l WETWELL TEMPERATURE i l I '
^
I l Figure 42 l
I Figure 5.5-3 I 700. 0,,-- . I-i : D l 500.0 l c : ' i2 : t I < E g 300.0 : r l N ! I l f 100.0: I b l l 0.0 2.0 4.0 6.0 8.0 TIME (SECONOS) X 1000 I I I GGNS BASE CASE SORY I CONTAINMENT TEMPERATURE l I l I I
, m .e I
i I z1 - s.s-4
- I g 40.0 ,
l
- 30.0 ll h ~
- e .
l I w - i 0 E 20.0 7
' /
'. /
l I .
/
/
/
+
E 10.0 5 0.0 ' 2.0 4.0 6.0 a.0 TIME (SECONOS) X 1000 l GCNS BASE CASE S0RV l . DRYWELL PRESSURE I I --_ -- -
I Figure 5.5-5 I 40.0 . . . - .. . I :
- 30.0 5 '
l $ g' N .
~
e '
\
? 20.0 l _
f7 I i A ,
, I 10.0 0.0 2.0 4.0 6.0 8.0 l TIME (SECONDS) X 1000 I
GCNS BASE CASE SORV g WETWELL PRESSURE g I - Figure 45 I ._ _ -
- I -
Figure 5.5-6 .I , l ,I . 40.0 . . . . I h b
- I
- 30.0 i
'l $ : ' s . I = a . I 20 E 20.0 f l : g I b I d 10.0 0.0 2.0 4.0 6.0 8.0 I TIME (SECONOS) X 1000 l CGNS BASE CASE SORV l CONTAINMENT PRESSURE I I I Figure 46 I
I Figure 5.5-7 , I I 5.0, .. l I* - .
! / li I ! *' ! / 1 ilJ 1.0 , ' ii .
l -1.0 ',
. i l,
IgW ' l - o g E ; _
~
I -s.0! 0.0 2.0 4.0 6.0 8.0 l TIME (SECONDS) X 1000 I GGNS BASE CASE SORV E DIFTERENTIAL PRESSURE I
~
!I I I o I - -- - - -
I Figure 5.5-8 I ' O.20 .
.l -
I : - m I a 0.is
' \
s< : N - I $ B . l 3 0.18 ,
- 0. 17 0.00 ?.00 4.00 6.00 8.00 g
TIME (SECONOS) X 1000
- GGNS BASE CASE SORV I
DRYWELL 02 CAS CONCENTRATION I I Figure 48 l l I
I! I m . m s.s-s e
?
l 0.30 , P D l5px 0.20
'). ,, -
t Iy O b Iso 0.10 I :
' (
l 0.00I- 6.00 8.00 0.00 2.00 4.00 ' 1 TIME (SECONOS) X 1000 g GGNS BASE CASE SORV 02 GAS CONCENTRATION WETWELL I I I n I e 4e I
l l Figure 5.5-10 0.30 . 1 l E 0.20 ~ 3 i A u
- 5 '. l w .
h -
@ 0.10 ,
g ,
'L6 t 9 l A a 2. k A t t t. E $ *,A Hg_Q))Q Q '=
^* ^ **
y
' O.00 2.00 4.00 6.00 8.00 TIME (SECONDS) X 1000 GGNS BASE CASE SORV CONTAINMENT 02 GAS CONCENTRATION Figure 50
f. I i Figure 5.5-11 { i l 0.74, P I ' 'N '
/ .
I! ! / I l! 0.72 i
/ '
lr3o> . ( - l : l - 0.70' Ltt
? 00 4.00 6.00 8.00 TIME (SECONDS) X 1000 g
GGNS BASE CASE SORV l DRYWELL N2 CAS CONCENTRATION I I l I l Figure 51 I
i Figure 5.5-12 iI I 0.80 j - l - i z : o , i l E ' \ N IE l 0.60 ' 1
, I E :
l
, 3 -
!I S : iI : I 0.40 8.00 2.00 4.00 6.00 l 0.00 TIME (SECONOS) X 1000 I GGNS BASE CASE SORV I N2 GAS CONCENTRATION WETWELL g I I I Figure 52 I
I Figure 5.5-13 1 0.80, . . I I j
) s l
Iap : i i. Iym i
~
0.60 . Ir3 ! N -
. t I
0.40 0.00 2.00 4.00 6.00 8.00 TIME (SECONDS) X 1000 I GCNS BASE CASE SORV I CONTAINMENT N2 GAS CONCENTRATION I ~ I I I Figure 53 1 I
.o I , Figure 5.5-14 I
0.02 -~ - b
'I :
I = : I $ e 0.01 7 g ra I 9 l I '
^
0.00 6.00 8.00 I 0.00 2.00 4.00 TIME (SECONOS) X 1000 I GGNS BASE CASE SORV jg H2 GAS CONCENTRATION ORYWELL g I 'I lI Figure 54 I
I Figure 5.5-15 I - l 0.10 . l l l. I 0.08 : {dl, g J raum, g$y 0.06 i Era g-E : i==- l w : essaiiii m 0.04 I yE j
- I '
O.02
- I {ll'1 ~ ~'I I 'l' :
/
l I 0.00' 2 - --- l-O.00 2.00 4.00 6.00 8.00 I TIME (SECONDS) X 1000 l . GCNS BASE CASE SORV l WETWEl.L H2 GAS CONCENTRATION I I I Figure 55 I
I I I 0.08; I : A A g Q# d 1 O.06 , l E 5 0.04 , I s 0.02 ! g ! l g
- 0. { 00 2.00 m )
4.00 6.00 8.00 l TIME (SECONOS) X 1000 GGNS BASE CASE SORY H2 GAS CONCENTRATION CONTAINMENT , lI I ' lI
, Figure 56 k
l I ' Figure 5.5-17 t I 0.11, . . . . . . . .. . j f: I - D v : I 0.09 I !* \ ' I ! \ ,
~
I !
\ !
- a I
0.07' -- O.00 2.00 4.00 6.00 8.00 l TIME (SECONDS) X 1000 I GCNS BASE CASE SORV I DRYWELL STEAM GAS CONCENTRATION I I I I . s> I - i
1 I rigure s.5-18 0.50 I I - z ! l @ y 0.30 11 I i y i 4 I 3 I 0.10 ; I !
- \(l
- h. 0.00 2.00 4.00 6.00 8.00 l TIME (SECONDS) X 1000 l l.
g CGNS BASE CASE SORV l WETWEU. STEAM CAS CONCENTRATION I I I nn se It _
I Figure 5.5-19 g 0.50, .
- L' l
E l E 0.30 A i E : :
- l w -
Z 3 o . i - I 0.10 , 'l 0.00 2.00 4.00 6.00 8.00 TIME (SECONOS) X 1000 I I CCNS BASE CASE SORV CONTAINMENT STEAM CAS CONCENTRATION I I I Figure 59 I I .. . -
I ,1, _ e.e . g g 300.0 . .
/
I :
-v ^*
y I c . w .
,200.0 I h e
z N - I ! I : 100.0 - - l 0.0 2.0 4.0 TIME (SECONDS) X 1000 6.0 8.0 I GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL ORAW DOWN DRYWELL TEMPERATURE I I I Figure 136 I I _
~
rigure 5.s-21 g I 2000.0 I : I - l l g :
= : ,
g h 1000.0 1
\
I o g - I I .
. . s.
0.0 6.0 8.0 0.0 2.0 4.0 l TIME (SECONOS) X 1000 I GCNS BASE CASE ORYWEtt BREAK SUPPRESSION POOL DRAW 00WN I WETWELL TEMPERATURE I I I Figure 137 I
Figure 5.5-22 lf
- l. 200.0
- l. 1
, 11 I
l L g E
! $ 100,.0
! 5 "
of $ i W '
,i l' '0.0 2.0 4.0 TIME (SECONDS) X 1000 6.b 8.0 I
I GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL DRAW 00VN ! g, CONTAINMENT TEMPERATURE , E 4 I. l Il . Ia . _
1 Di g
,1e_ e . e-u g .
l 30.0 . . . 1 1 : I 2 [ ! g E 3 T\ I y (r 8
, x
/
I : I
! i I 10.0 0.0 2.0 4.0 6.0 8.0 TIME (SECONDS) X 1000 g GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL DRAW DOWN g ORYWELL PRESSURE I
I I ne g I
rigure 5.5-24 h l 30.0 ' ' ' I . j 1 ! , 1 I f. D N (H c- I l w g 20.0 L -[l ' a I e. x 7 ( . l : - l _l 0.0 2.0 4.0 6.0 8.0 TIME (SECONOS) X 1000 g GGNS BASE CASE ORYWELL BREAK SUPPRESSION P0OL DRAW 00WN I WETWELL PRESSURE I I . nn w I:
I I l 30.0 , I : g j :j I 3 '
,i f
- - p l
g 20.0 m - f LJ 8 . l E
.(
I . I : 10.0 I 0.0 2.0 4.0 TIME (SECONOS) X 1000 6.0 8.0 l GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL DRAW DOWN CONTAINMENT PRESSURE g I i l I I Figure 141 I ./
e I ngure s.s-26 5.0,
.l; I: -
u.i 3 l- .! V 3.0 , g F' )
!!, z .
l
- x :
ll
! {
to ; s ? iii lJ , i 5 ao i
~
l l l' s i ll;
- -1.0 0,Q 2.0 4.0 IIME (SECONDS) X 1000 6.0 8.0
'g GGNS BASE CASE DRYWELL BREAK l~ SUPPRESSION POOL DRAW DOWN l Olf7ERENTIAL PRESSURE I I o I;
"=
ll d
Figure 5.5-27 [ 0.20 5 1 5 { 0.10 { s l ~
/ ~
[
~
A W r 0.00 - 0,00 0,00 4,00 6.00 8.00 TIME (SECONOS) X 1000 { GGNS BASE CASE ORYWELL DREAK SUPPRESSION POOL DRAW 00WN ORYWELL 02 GAS CONCENTRATION i { Figure 143 (
d I I . l 0.30 I 5 0.20 ~ I E m c-I i . ' l , 3 y -0.10 l I I
- .0.00 .
g 0.00 2.00 4.00 6.00 8.00 TIME (SECONDS) X 1000 l GGNS BASE CASE ORYVELL BREAK SUPPRESSION POOL DRAW DOWN l WETWELL 02 GAS CONCENTRATION I I I un n g I
I I r1eere e s -2e l 0.30 I :
~
, I : E I 6 6 ' I O.20 m w - M E I e I ! I .0.10 0.00 2.00 4.00 6.00 8.00 TIME (SECONDS) X 1000 g CCNS BASE CASE ORYWELL BREAK SUPPRESSION POOL ORAW DOWN g CONTAINHENT 02 GAS CONCENTRATION I I I
-. u, I. _ _
r
I Figure 5.5-30
! 0.80. - ' ' '
I : I 0.80I i = e
! 0.40 -
I ie I 0.20 .
~ .
g
~
/
y . I 0.00' 0.00 2.00 4.00 6.00 8.00 TIME (SECONDS) X 1000 g GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL ORAW 00WN DRYWELL N2 CAS CONCENTRATION I . I E _ g nn - I b
-- a a -=
1 I I I o 8o I 5 0.70 l 5 l , . 2 I g 0.s0 1, , l : I i 0.50 - l 0.00 2.00 4.00 6.00 8.00 TIME (SECONOS) X 1000 g CGNS BASE CASE ORYWELL BREAK SUPPRESSION P0OL ORAW D0WN WETWELL N2 GAS g CONCENTRATION I I I - Figure 147 I- . j
I l Figure 5.5-32
- g 0.80 N
I I
- l z 9
i lI O.70 I r 2 I " I . 0.60 6.00 8JJO 0.00 2.00 4.00 I - TIME (SECONOS) X 1000 I GGNS BASE CASE ORYWEli BREAK SUPPRESSION P0OL DRAW 00VN g CONTAINMENT N2 CAS CONCENTRATION I I I - Figure 148 E.
l I Figure 5.5-33 I l 0.50, I D E I b
< 0.30 .
I w : I a - S [ j 0.10 ' i - l I 0.00 2.00 4.00 6.00 8.00 TIME (SECONOS) X 1000 g GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL DRAW 00VN DRYWELL g H2 GAS CONCENTRATION I I lI Figure 149 ,I I ;
a I rigure 5.5-34 I . 0.08 .- g U 0.06 0.04!
' i* .
1 H lll 8 H,,I H,I g. 0.02 !
) l
" 2 I- l 0.00; 6.00 8.00 2.00 4.00 0.00 g TIME (SECONDS) X 1000 I GCHS BASE CASE ORYWELL BREAK SUPPRESSION P0OL DRAW 00WN H2 CAS CONCENTRATION
- l WETWELL I,
I Figure 150 50 . - - _ _ _- .. . _
l Figure s.s-3s I 0.07. - f ' 0.05 ! i l 5 g y 0.03 ! I 3 : 7 l $ ! f I .
/
l 0.01 [ ' O.00 2.00 4.00 6.00 8.00 g TIME (SECONOS) X 1000 GCNS BASE CASE ORYWEli BREAK SUPPRESSION POOL ORAW 00VN CONTAINMENT H2 GAS CONCENTRATION I I 'I g .m I
Figure 5.5-36 i 1.00 - - i f N ' I.l . N 0.80: N l: \ l S
?3 0.60 g
- l. $ \
l N E 0.40
\
S S ll P 0.20 ' l3 I .
^
ll' i 0.00' 0.00 2.00 4.00 6.00 8.00 j l > t TIME (SECONDS) X 1000 r GGNS BASE CASE DRYWELL BREAK l SUPPRESSION POOL DRAW 00VN DRYWELL STEAM GAS CONCENTRATION I I: i Figure 152 Ii - - - _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ . _ = .
I l ngue s.s-n i 0.40, . . . . L l ! . Ad 0.20 r' . I i
- i I .
u t l 0.00' - O.00 2.00 4.00 6.00 8.00 J TIME (SECONDS) X 1000 l G"S BASE CASE ORYWELL BREAK SUPPRESSION POOL ORAW 00VN l WETWELL STEAM GAS CONCENTRATION lI l - I
, m . 1s3 l .
I '
"e-e s s-a I
I 0.30 . I : I l 5 e 0.20 , i 3o 0.10 , 7 , I : . J !, 0.00 2.00 4.00 6.00 8.00 0.00 l TIME (SECONOS) X 1000 I GGNS BASE CASE ORYVEli 8REAK SUPPRESSION P00L DRAW 00WN g CONTAINMENT STEAM CAS CONCENEATION
- I l
l I n= 'I I 4
~~
I Figure 5.5-39 g 800.0. .. l 600.0 ! l 6 f E g a: - g 400.0 : : I : I l ( 200.0 : ( g 7.0 9.0 11.0 13.0 TIME (SECONDS) X 1000 I GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL DRAW 00VN ORYWELL TEMPERATURE t Figure 155 I
1 I Figure 5.5-40 I . 3000.0 . .
'l i . .
- I
~
i 2000.0
- l 6 .
u i g E . IE I w !l W 1000.0 I '. l '
, 0.0
'l 7.0 9.0 11.0 13.0 TIME (SECONOS) X 1000 I GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL ORAW 00WN WETWELL TEMPERATURE I I Figure 156 i
't
,I Figure 5.5-41 !,I i l 900.0. i
- I i :
700.0 : . E ,
@ 500.0 l
= :
E ! I z - W : 1 l . ll 300.0 j k ' i LL L L l 100.0'-
- ^
7.0 9.0 11.0 13.0 I TIME (SECONDS) X 1000 l CGNS BASE CASE ORYWELL BREAK l SUPPRESSION POOL ORAW DOWN CONTAINMENT TEMPERATURE I I Figure 157 I
I , rigure s.s-u 40.0 - I
~
I s. . b w 30.0 s . l 8 u a. 4 l l Y i-y 'g L
\\
20.0 13.0 7.0 9.0 11.0 l TIME (SECONDS) X 1000 I CGNS BASE CASE ORYWELL BREAK I SUPPRESSION POOL DRAW DOWN ORYWELL PRESSURE , I I I Figure 158 I
I Figure 5.5-43 I , i I 50.0, I a i I w s 30.0 , l 8 E
\
I l L z
\.
I i I 10.0 7.0 9.0 11.0 13.0 g TIME (SECONDS) X 1000 I GCNS BASE CASE ORYWELL BREAK SUPPRESSION POOL DRAW DOWN WETWELL PRESSURE I I I Figure 159 I
I
"*""-4'
- I l
50.0 ,- - . D D D .I :. e : l w 30.0 ,
- i. s, v
I W '
- : : x L, t_ \
K ci ~ w %
- 10. 11.0 13.0 0 9.0 TIME (SECONDS) X 1000
; l GGNS BASE CASE ORYVELL BREAK SUPPRESSION P0OL DRAW DOWN ,
l . CONTAINMENT PRESSURE
'I
- g Figure 160 4
_.L_ .__ _. _ __ ._.._. _ _ _ . __ _ -.__.._._
_ q I ' i l Figure 5.5-45 0 1 I 10.0 , w . E - I= ; v r( ( < G 0.0 -s Iou - 7 l Z . I -10.0 y , l W c I l
-20.0 ^ ^
7.0 9.0 11.0 13.0 TIME (SECONOS) X 1000
! GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL DRAW DOWN I DIITERENTIAL PRESSURE I
~
I I l ,12 1e1
- I
I Figure 5.5-46 !I ll - 0.05 [ ]. ,k 4- : - l . /
/ '
~
! 5<: 0.03
/ .
l l
/ y i
[ I W : l-l 3 o
)
i ! l 1 i 0.01 '. 'I i
^
11.00 C.00 i 7.00 9.00 TIME (SECONos) X 1000 I 'I
.I GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL ORAW 00WN 02 GAS CONCENTRATION ORYWEli I
I I Figure 162
r.
~ Figure 5.5-47 I 0.20 I i -
I= : 5 . I ' I O.10 ( f Iy ' s
$ p -
[ I : I 0.00
- ^^ ^
S.00 11.00 13.00 7.00 g TIME (SECONOS) X 1000 I GGNS BASE CASE ORYWELL BREAK c SUPPRESSION POOL DRAW 00VN WETWELL 02 r.AS CONCENTRATION 1 I I - Figure 163
.I Figure 5.5-48 i
0.12. - Il
! [h 7 L
- a i
/ (
g .
)
- 0. 10: ~
l' 5 . l P u I E 4 l y 3 o 0.08 ; l i I t, - , I 0.06 '.
/
i m 13.00 9.00 11.00 7.00 TIME (SECONOS) X 1000 l :
- . g GGNS BASE CASE ORYWELL BREAK SUPPRESSION P0OL ORAW 00WN ll 02 GAS CONCENTRATION CONTAINMENT I
I Figure 164
\tl I Figure 5.5-49 l 0.50, -
. r
- /
I ! [ .
'l 0.30 l E ! /
l d i
- / .
i / I : l 0.10 7.00 9.00 11.00 13.00 g TIME (SECONDS) X 1000 GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL DRAW DOWN ORYWELL N2 GAS CONCENTRATION l I I . I
I I 0.70 . . I - I B I v 0.60 i . O , l ; 0'50 11.00 13.00 7.00 9.00 l TIME (SECONOS) X 1000 GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL DRAW 00WN l N2 GAS CONCENTRATION WETWELL I . I rigure 166 l
y-I Figure 5.5-51 I I 0.70 -- l l . W f I = E I E 0.60 l E so lI . 0.50 13.00 7.00 9.00 11.00 I TIME (SECONOS) X 1000 ,I i GGNS BASE CASE ORYWELL BREAK il CONTAINMENT SUPPRESSION POOL DRAW 00WN N2 GAS CONCENTRATION
~
t 1I I i I Figure 167 5 __ _ _ . _ .
Figure 5.5-52
.I I y v w w y y 0.50.
l v
) ,
- s .
\ :
N
\
? 5 i I
- a E
0.30 1 l E . 3 h-m 13.00 O.10l 9.00 11.00
. 7.00 j
TIME (SECONOS) X 1000 GGNS BASE CASE ORYWEl.L BREAK SUPPRESSION POOL DRAW 00VN H2 CAS CONCENTRATION DRYWELL Figure 168 s
- - - ~ . _ _ _ . _ __ _ _ _ _ - - - . . _ _ _
\._
I Figure 5.5-53 { I i l 0.20 - l I : I IE < l m
- 0. 10 f
> ( .
I ' 4 I . c-l i 0.00 11.00 13.00 7.00 9.00 TIME (SECONOS) X 1000 g GGNS BASE CASE ORYVELL BREAK SUPPRESSION P0OL DRAW 00WN H2 GAS CONCENTRATION WETVELL I I I Figure 169 I
E Figure 5.5-54 I l l 0.08.
! A d .
I : i /
- / -
' I O.06: 5 i - l A a I y 0.04 ; I 3 e ; I l g 0.0h j - i g-I 7. 0 9.00 11.00 13.00 TIME (SECONDS) X 1000 g GGNS BASE CASE ORYVELL BREAK SUPPRESSION POOL DRAW 00VN H2 GAS CONCENTRATION CONTAINMENT I I I I Figure 170 I -
I Figure 5.5-55 I l 0.60 ' V
~
I
~
5 g : E ' 4: I 6 O.40 ', 3 I E :
! Y l :
N _ 0.20' 13.00 7.00 9.00 11.00 l TIME (SECONOS) X 1000 I GGNS BASE CASE ORYVELL BREAK SUPPRESSION POOL ORAW 00VN g ORYWELL STEAM GAS CONCENTRATION I
- I I
I Figure 171 I A
I ,1, _ s.s-se
- I l
0.50, : l , 5 . l C : l W , l g . i 0.30 ', l Il y 3
\: '
!l l 5 :
1 P w 4i 0.10' - - l 7.00 9.00 11.00 TIME (SECONDS) X 1000 13.00 g GGNS BASE CASE ORYWELL BREAK I SUPPRESSION POOL DRAW DOWN f WETVELL STEAM GAS CONCENTRATION c l Figure 172 , v ^m . er _
m I Figure 5.5-57 I 0.40 , I : I ~ E 0.30 ' IE< : E I Ii 0.20 i
\3 i
l : W h J -
'l 0.10 - -
7.00 9.00 11.00 13.00 TIME (SECONOS) X 1000 g
! GGNS BASE CASE ORYWELL BREAK SUPPRESSION POOL ORAW DOWN i
l I CONTAINMENT STEAM GAS CONCENTRATION I I I ! Figure 173 I
I DRYWELL EQUIPMENT SURVIVABILITY LIST TABLE 5.6-1 EQUIPMENT IDENTIFICATION Rx CEN-EQUIPMENT AZI- TERLINE NUMBER + DESCRIPTION QUALIFICATION FUNCTION ELEVATION MUTH DISTANCE TEMP (F) DURATION MANUF. MODEL 1B21F0041A Atttomatic Depressuri- RPV Pressure 636' 5" 51 20' 355 3 hrs Dikkers C471-6/125.04 zation System Valve Relief / ADS IB21F0041B " " " " 636' 5" 277 26' " " " " 1B21F0041E * *
- 636' 5" "
31 21' * *
- IB21F0041F " " " "
636' 5" 289 26' " " " " 1B21F0047D " " " " IB21F0047H " * *
- 636' 5" 308 20 " * *
- IB21F0051C * * * "
636' 5" 322 21' * * *
" 636' 5" 88 25' " " "
1B21F0051G " " " 636' 5" " IB21F0410A " 71 26' * "
- Automatic Depressuri-
- Location Same as Valves zation System Valve Qualification Seitz 6A33 Solenoid in Progress IB21F0410B " = a a = = = = = = =
IB21F0411A * = = = = = = = = = = IB21F0411B ' " = = = = = = = = = = 1821F0414A = = = = = = = = = = = IB21F0414B " = = - = = = = = = = IB21F0415A " " = " = = = = = = = IB21F0415B " * * = = = = = = = = IB21F0422A * * = = = = = = = = = IB21F0422B " " = = = = = = = = = IB21F0425A * " " = = = = = = = = IBSIF0425B " " = " = = = = = = = IB21F0442A " " = " = = = = = = = IB21F0442B " " = = = = o = = = = IB21F0444A * * * = = = = = = = . IB21F0444B " " = = = = = = = = = 1D23N0100A Drywell RTD Drywell Temp. 642' 315 17' 485 3 hrs Weed 611 Monitoring ID23N0100B " " = " 642' 135 16' " " " " 1D23N0110A " " " " "
- 620' 6" 308 36' 6" " " "
1D23N0110B * * " 620' 6" " l
" 145 36' 6" " " "
ID23N0120A - " " " 599' 9" 308 36' 6" " * " * { ID23N0120B " " " " 599' 9" 150 * *
- 36' 6"
- l 1
m m m m m m m m M M M M M M M
i Page 2 DRYWELL EQUIPMENT SURVIVABILITY LIST TABLE 5.6-1 (Cont.) EQUIPMENT Rx CEN-IDENTIFICATION EQUIPMENT AZI- TERLINE QUALIFICATION NUMBER DESCRIPTION FUNCTION ELEVATION MUTH DISTANCE TEMP ( F ) DURATION MANUF. MODEL 1M56S008 Ilydrogen Ignition 11ydrogen 629' l-1/2" 12 36' 6" 345 3 hrs Power 6043 System Ignition Systems
" " " " " " =
IM56S009 637' 0" 41 36' 6" IM56S010 " 636' 3-1/2" 90 36' 6" " " IM56S011
- 636' 7" 137 36' 6" "
- IM56S012 "
632' 3" 180 36' 6* * * = =
" * * * " = "
IM56S013 631' 5" 221 36' 6" 1H56S014 " 636' 10" 273 36' 6" * *
* * * = " " =
IM56S015 630' 9-1/2" 322 36' 6" IM56S016 660' 0" 0 31' 6" IM56S017 " 659' B" 57 29' 6" * " " " 1M56S018 659' 8" 114 30' 0" " = IM56S019 " " 659' 8" 172 30' 0* *
- e "
" " " " = " "
IM56S020 659' 8" 225 28' 0"
* * * " " = "
1M56S021 660' 0* 280 30' 0" 1M56S022
- 660' 0" 317 31' 0" " "
1M56S102 670' 0" 350 13' 0" *
- IM56S103 670' 0" 4 13' 0" Control Cable and Drywell 346 Rockbestos Firewall III Small Power Cable (Various Locations)
Instrument Cable 385 Brand-Rex 16 & 20 AWG Drywell Personnel 603' 1" 105 36' 6" In W. J. Wooley Airlock Seal Progress Drywell Equipment 605' 227 36' 6" = Haech Sea 1 I i
CONTAINMENT EQUIPMENT SURVIVABILITY LIST TABLE 5.6-2 EQUIPMENT Rx CEN-IDENTIFICATION EQUIPMENT AZI- TERLINE QUALIFICATION NUMBER DESCRIPTION FUNCTION ELEVATION HUTH DISTANCE TEMP (F) DURATION MANUF. MODEL ID23NO130A Containment RTD Containment 689' 0" 272 60' 485 " Weed 611 Temperature Monitoring " " 1D23NO130D
" 720' 0" 95 60' 1D23NO140A " 664' 0" 45 60' 1D23NO140B
" 664' 0" 210 60' 1D23NO150A 642' 0" 55 60' 1D23NO150B " 642' 0" 250 60' 1D23NOl60A
- 599' 9" 67 60' *
- ID23NOl60B "
599' 9" 250 60' lE12F0028A Containment Spray Containment Spray 643' 6" 37 48' 6" 340 Limitorque SMB Isolation Valve (MO) " " " * *
- lE12F0028B " 643' 9" 335 42' 9"
" " SMB lE12F0042A RHR LPCI Inboard Low Pressure 624' 0" 41 44' 0" Isolation Valve (MO) Coolant Injection " "
IE12F0042B
*
- 620' 0" 315 55' 0" lE12F0537A Containment Spray Containment Spray 689' 0" 40 58' 0" Isolation Valve (MO) " " " "
lE12F0537B 689' 0" 320 58' 0"
" Henry Pratt NRS 1M16F0010A Drywell Vacuum Drywell Isolation 652' 325 36' 6" 250 Relief System Butterfly Valve " "
IM16F0010B * * "
- 652' 222 36' 6" 250 1M16F0020A " "
Drywell Isolation 652' 324 36' 6" 250 " GPE Controls LD240-339 Check Valve " *
- IM16F0020B " " " "
652' 225 36' 6" 250 IM17F0010 Containment Vacuum Containment Vacuum 664' 58 60' 250 *
- LD240-337 a Relief System Releif Check Valve * " "
IM17F0020 " " " " 664' 150 60' 250 " 1M17F0030 " " " " 664' 302 60' 250 " 1M17F0040 * * " " 664' 315 60' 250 4
Page 2 CONTAINMENT EQUIPMENT SURVIVADILITY LIST TABLE 5.6-2 (Cont'd) EQUIPMENT Rx CEN-IDENTIFICATION EQUIPMENT AZI- TERLINE QUALIFICATION NUMBER DESCRIPTION FUNCTION ELEVATION MUTH DISTANCE TEMP (F) DURATION MANUF. MODEL 1M51C0001A Hydrogen Mixing Hydrogen Mixing 664' 300 24' 192 2 days Turbonetics SC-6 Compressor and Compressor Motor Reliance Type P Motor IM51C0001B " *
- 664' 245 25' 192 IM51D0001A Hydrogen Recombiner Removal of Hydro 664 304 37' 1700-1750 21 days Westinghouse Model A gen by Hydrogen (Heater and Oxygen Element)
Recombination IM51D001B " " 664' 236 37' IM51F0010A Hydrogen Mixing Isolation valve 670' 309 25' 340 3 hrs Limitorque SMB-00-5 Compressor Iso- for Drywell lation Valve (MO) Purge Compressor IM51F0010B " " " 670' 245 20' 340 1M51F0501A Hydrogen Mixing Check Valve for 664' 305 25' 350 = TRW Mission K15ACEFV73 Compressor for Drywell Purge l Check Valve Compressor " " IM51F0501B " " " 250 21' 350
- 1M56S001 Hydrogen Igniter Hydrogen Ignitioq 613' 4" 355 49' 0" 345 3 hrs Power Sys. 6043 System Division IM565002 613' 4" 5 51' 0"
, IM56S003 " " 619' 6" 63 51' 8" i IM565004 " " " 619' 6" 89 52' 0" i IM56S005 " " 664' 0" 34 57' 0" 1M56S006 *
- 689' 0" 34 52' 0*
lH56S023 " " 619' 6" 34 52' 0"
- IM565024 " *
- 619' 6" 118 51' 8" IM56S025 * " * "
619' 6" 152 51' 0" IM56S026 " " " " 619' 6" 186 52' 0" 1M56S027 * " 619' 6" 221 51' 8" IM56S028 * " *
- 619' 6" 255 51' 4" IM56S029 " " " "
619' 6" 289 52' 0" M M M M M M M M M m
Page 3 CONTAINMENT EQUIPMENT SURVIVABILITY LIST TABLE 5.6-2 (Cont'd) EQUIPMENT Rx CEN-IDENTIFICATION EQUIPMENT AZI- TERLINE QUALIFICATION NUMBER DESCRIPTION FUNCTION ELEVATION MUTH DISTANCE TEMP (F) DURATION MANUF. MODEL 1M56S030 Uydrogen Igniter Hydrogen Ignition 619' 6" 322 51' 11" 345 3 hrs Power Systems 6043 System Division IM56S031 638' 0" 358 41' 6" " " " 1M56S032 640' 0" 155 46' 0" IM56S033 640' 0" 186 46' 0" " " " " 1M56S034 640' 0" 324 53' 6" 1M56S035 640' 4-3/4" 61 51' 6" IM56S036 " 640' 5-1/2" 118 51' 6" " 1M56S037 640' 5" 227 46' 0" IM56S038 639' 4" 260 54' 0" , IM56S039 651' 1" 286 41' 6" IM56SO40 647' 4" 2 41' 6" 1M56SO41 650' 6-3/4" 41 50' 6" 1M56SO42 650' 6" 87 49' 0" 1M56SO43 651' 0" 101 49' 0" IM56S044 660' 0" 86 44' 6" 1M56SO45 660' 6" 95 48" 6" 1M56SO46 664' 0" 54 51' 0" 1H56SO47 665' e" 114 52' 0" 1M56SO48 662' 6" 147 53' 0" 1M56SO49 662' 7-3/4" 218 51' 0" 1M56S050 664' 7" 251 49' 6" 1M56S051 661' 6" 289 50' 0" 1M56S052 661' 6" 324 49' 6" 1M56S053 669' 6" 0 54' 6" IM5 6S 054 684' 9" 355 52' 6" 1M56S055 686' 0" 75 48' 0" 1M56S 056 686' 0" 85 47' 0" IM56S057 686' 0" 95 47' 0" 1M56S058 686' 0" 105 48' 0" 1M56S059 686' 0" 75 35' 0" IM56S060 686' 0" 105 35' 0" 1M56S061 689' 6" 45 48' 0" M M M M M M
Page 4 CONTAINMENT EQUIPMENT SURVIVABILITY LIST TABLE 5.6-2 (Cont'd) EQUIPMENT Rx CEN-IDENTIFICATION EQUIPMENT AZI- TERLINE QUALIFICATION NUMBER DESCRIPTION FUNCTION ELEVATION MUTH DISTANCE TEMP (F) DURATION MANUF. MODEL 1M56.062 Hydrogen Igniter Hydrogen Ignition 689' 6" 130 41' 0" 345 3 hrs Power Systems 6043 System Division IM56S063 *
- 689' 6" 229 48' 0" " " *
- IM56S064 *
- 689' 6" 252 43' 6" " " " "
1M56S065 " 689' 6" 289 43' 0" " " " " 1H56S 06 6 689' 6" 310 48' 6" 1M56S 067 715' 6" 359 58' 9* 1M56S 068 715' 6" 27 58' 9" 1M56S 069 715' 6" 62 58' 9"
" " = " * "
1M56S 070 715' 6" 87 58' 9" " " 1M56S 071 715' 6" 119 58' 9" 1M56S 072 715' 6" 151 58' 9" 1M56S 073 715' 6" 178 58' 9" 1M56S 07 4 715' 6" 209 58' 9" IM56S 075 " 715' 6" 241 58' 9" 1M56S 076 715' 6" 273 58' 9" IMSGS077 * " " " " #" " " 715' 6" 300 58' 9" 1M50S078 " 715' 6" 331 58' 9" " " IM56S 079 745' 6" 359 48' 0" 1M5 6S 08 0 745' 6" 34 48' 0" 1M5 6S 081 745' 6" 72 48' 0" IM56S082 * * " " 745' 6" 102 48' 0" " " " " 1M5 6S 08 3 745' 6" 143 48' 0"
- IH5 6S 08 4 745' 6" 180 48' 0" 1M56S 08 5 745' 6" 216 48' 0" 1M56S 08 6 745' 6" 252 48' 0" 1M5 6S 087 745' 6" 287 48' 0" 1M56S 088 745' 6" 324 48' 0" 1M56S089 757' 0" 0 l' 0" 1M5 6S O90 757' 0* 180 l' 0" 1M56S O91 645' 7" 168 60' 0*
1M5 6S O92 645' 0" 172 58' 0" ! 1M56S O 9 3 613' 4" 7 44' 0" l l l
Page 5 CONTAINMENT EQUIPMENT SURVIVADILITY LIST TABLE 5.6-2 (Cont'd) EQUIPMENT Rx CEN-IDENTIFICATION EQUIPMENT AZI- TERLINE QUALIFICATION NUMBER DESCRIPTION FUNCTION ELEVATION MUTil DISTANCE TEMPtF) DURATION NJR NU P. MODEL 1M56-094 Hydrogen Igniter Hydrogen 612' 5" 13 42' 8" 345 3 hrs Power Systems 6043 System Ignition Division 1H560095 * *
- 612' 6" 344 42' 6" " " "
- IM56SO96 *
- 612' 3" 351 43' 6" " * *
- IM56S097 " " "
638' 8* 289 49' 6" " " " " 1M56SC98 *
- 685' 6" 342 53' 0" " * "
- IM56SO99 " "
6 8 'a ' 6" 17 50' 6" " " " " 1M56S100 " " 6'46' 0" 75 25' 0" " " " " 1M56S101 * *
- r,8 6 ' 0" 105 25' 0" " " *
- 1R7250001 Electrical Penetrations Containment 659' 0* 221 60' 340 3 hrs Westinghouse WX33328 Boundary IR72S0002 " " "
659' 0" 228 * * * " WX33328 1R725D003 *
- 656' 3" 221 * * *
- WX33329 1R72S0004 " "
657' 1-1/2' 248 * * " " WX33329 1R72S0005 *
- 656' 3" 228 * " " "
WX33330 1R72S0006 * * " 657' l-1/2" 242 * * *
- WX33331 1R72S0007 * *
- 651' 6" 221 * * " "
WX33332 1R7250008 " " " 649' 9" 221 " " " " WX33333 1R72S0009 " " "
- 651' 6" 248 " " "
WX33332 1R7250010 " *
- 649' 9" 248 " " " "
WX33333 1R72S00ll " " " 657' l-1/2" 235 " " " " WX33334 1R72S0012 " " " 651' 6" 228 * " " " WX33335 1R72S0013 * * " 649' 9" 228 " " " " WX33333 1R7250014 " " " 651' 6" 242 " " " " WX33335 1R72S0015 " " " 649' 9" 242 " " " " WX33333 1R7250016 " " " 643' 3" 221 " " " " WX33336 1R72S0017 * "
- 641' 6" 221 " " " "
WX33337 1R7250018 " " " 643' 3" 228 " " " " WX33338 1972S0019 * " " 641' 6" 228 * * " " WX33339 1R72S0020 " " 643' 3* 248 * " *
- WX33336 1R72S0021 " " "
641' 6" 241 * * * " WX33363 1R72S0022 * *
- 643' 3" 242 * * " "
WX33340 1R72S0023 " *
- 641' 6" 248 * * *
- WX33341 E E E E
Page 6 CONTAINMENT EQUIPMENT SURVIVABILITY LIST i TABLE 5.6-2 (Cont'd) EQUIPMENT Rx CEN-IDENTIFICATION EQUIPMENT AZI- TERLINE OUALIFICATION NUMIER DESCRIPTION FUNCTION ELEVATION MUTH DISTANCE TEMP (F) DURATION MANUF. MODEL ~ 1R72S0024 Electrical Penetrations Containment 643' 3" 235 60' 340 3 hrs Westinghouse WX33342 Boundary 4 1R72S0025 651' 6" 235 WX33337 1R72S0026 638' 4" 221 WX33343 1R7250027 638' 4" 228 ". WX33344 1R7250028 641' 6" 223 WX33345 1R72S0029 656' 3" 223 W-34147
. 1R72S0030 643' 3" 223 W34488 1R72S0031 649' 9" 223 W-34489 1R7250033 649' 9" 235 W-34490 1R72S0035 641' 6" 242 W-34491 1R72S0036 649' 9" 241 W-34492 1R72S0038 " " "
651' 6" 241 " W-34493 Upper Personal Airlock 692' 10" 225 60' Qualification J. Hooley Seals In Progress Lower Personal Airlock 603' 1" 241 Seals Equipment Hatch Seals 629' 6" 133 Terminal and Fuse Containment 346 3 hrs Buchanan NDQ, NQO, Block Assemblies (Various Locations) NQO-361 Control Cable and.
- 346 Rockbestos Firewall III Small Power Cable' Instrument Cable 385 Brand-Rex .16 and 20 AWG Pressure / Level /DP
- 318 Rosemont 1153 Transmitters
~
Pressure / Level /DP
- 232 Rosemont 1152 Transmitter i
TABLE 5.6-3 GRAND GULF NUCLEAR STATION DRYWEI.L EQUIPHENT REQUIRED TO SURVIVE A IlYDROGEN BURN Dsst. from Qualification Equipment Center Line or Design Identification Function Elevation Azimuth of Reactor Temperature Duration Combustible Gas Control System (CCCS) 3 llours E61-D106 'llydrogen Igniters 146'-3 7/8" 0 22'-10" 330*F Ilydrogen Igniters 145'-7" 63 29'-3" 330*F 3 Hours E61-D107 E61-D108 Ilydrogen Igniters 146'-2" 120 29'-8" 330*F 3 llours E61-D109 Ilydrogen Igniters 147'-l" 180 26'-3" 330*F 3 llours E61-Dilo liydrogen .'gniters 148'-7" 240 29'-l 1/2" 330*F 3 llours E61-Dill llydrogen Igniters 145'-7" 313 25'-l I/4" 330*F 3 llours
- E61-Dil2 Ilydrogen Igniters 160'-7 7/8" 0 27'-3 3/8" 330*F 3 llours 160'-11 3/4" 60 29'-8 3/4" 330*F 3 Ilours E61-Dil3 Ilydrogen Igniters E61-Dil4 Ilydrogen Igni t ers 160'-4" 135 27'-0 3/8" 330*F 3 llours E61-Dil5 Ilydrogen Igniters 160'-11 1/2" 180 26'-10" 330*F 3 llours E61-Dil6 Ilydrogen Igniters 160'-6" 232 26'-1" 330*F 3 llours E61-DlI7 Ilydroger. Igni t ers 160'-6" 324 26'-4 5/8" 330*F 3 llours l
E61-Dil8 Ilydrogen Igniters 179'-0" 0 26'-4 5/8" 330*F 3 llours E61-Dil9 Ilydrogen Igniters 179'-0" 65 26'-3 3/4" 330*F 3 llours 179'-0" 125 26'-3 3/4" 330*F 3 llours E61-D120 llydrogen Igniters E61-D121 Ilydrogen Igniters 179'-0" 185 26'-3 3/4" 330*F 3 Ilours E61-D122 Ilydrogen Igniters 179'-0" 245 26'-3 3/4" 330*F 3 llours E61-D123 Ilydrogen Igniters 179'-0" 305 26'-3 3/4" 330*F 3 llours Transformers for Igniters Respective Locations 400*F(l) -- Nuclear lloiler System (NBS) 4 Days B21-F047A ADS (A.D.) 154'-0" 34 22'-0" 349"F 154'-0" 315 21'-0" 349*F 4 Days B21-F041D ADS (A.O.) 4 Days B21-F047L ADS (A.O.) 154'-0" 53 27'-6" 349'F 154'-0" 288 26'-6" 349'F 4 Days B21-F041F ADS (A.D.) 4 Days B21-F041K ADS (A.O.) 154'-0" 304 27'-0" 349'F 154'-0" 45 22'-0" 349"F 4 Days B21-F051A ADS (A.O.) 4 Days B21-F051B ADS (A.D.) 154'-0" 272 25'-6" 349'F 154'-0" 77 26'-0" 349 F 4 Days B21-F051C ADS (A.O.) E E M M M M M
TABLE 5.6-3 (Continued) GRAND GULF NUCLEAR STATION DRYWELL EQUIPHENT REQlllRED TO SURVIVE A IlYDROGEN HllRN Dist. from Qualification Center Line or Design Equipment Duration l Function Elevation Azimuth of Reactor Temperature Identification Residual lleat Removal System Isolation Valve (H.0). 124'-7" 0 25'-0" 340*F -- l E12-F009 340*F -- Hotor Operator Same as Valve Position Indication Switches Same as Valve 340*F -- Area Radiatinn Honitoring System 6 Hours Radiation 161'-10" 0 36'-0" 340*F D21-HE-N048A 6 Ilours D21-HE-N048D Honitors 161'-10" 183 36'-0" 340*F Containment and Drywell Instrumen-Lation and Control System 340*F 6 Ilours Temperature Honitors 16I'-10" 40 36'-0" H71-TE-N008A H11-TE-N00811 161'-0" 250 36'-0" 340*F 6 liours 161'-0" 135 36'-0" 340*F 6 Hours H71-TE-N008C 340*F 6 Ilours if71-TE-N008D 161'-0" 310 36'-0" 94'-6" 55 10'-7" 340*F 6 liours H71-TE-N013A 6 Ilours H11-TE-Nol3B 94'-6" 225 10'-7" 340*F 94'-0" 112 10'-3" 340*F 6 Hours H7l-TE-N013C H71-TE-N013D 94'-6" 280 10'-7" 340*F 6 Ilours Containment flatches
,d Lock Drywell Personnel I.ock 124'-8" 60 36'-6" 330*F --
H23-Y005N 330*F -- H23-Y009 Drywell Equipment Hatch 117'-4" 220 36'-6" M M M m m m a m m m
TABLE 5.6-3 (Continued) GRAND GULF NUCLEAR STATION DRYWEl.I. EQUIPHENT REQUIRED TO SURVIVE A IlYDROGEN BURN D i .< t . from Qualification Center Line or Design Equipment Identification Function Elevation Azimuth of Reactor Temperature Duration Various Systems Power Cable (2) (2) (2) 346*F 3 Hours, 20 Minutes i 1 Control Cable (2) (2) (2) 346*F 3 llours, ) 20 Minutes d i instrument Cable (2) (2) (2) 340*F 6 Ilours Thermocoaple Ext. Wire (2) (2) (2) 340*F 5\ Ifours (2) (2) (2) 340*F 5 Hours Terminal Boxes and Blocks i 1 l 1 l l l l NOTES: A.O. = Air Operated H.O. = Hotor Operated (1) Underwriters Laboratory approved maximusi temperature for continuous operation at rate electrical load. (2) Specific routing will be evaluated on a case by case basis. M E M M M M M M M M M m M fM
TABLE 5.6-4 GRAND GULF NUCLEAR STATION CONTAINHENT EQUIPHENT (OllTSIDE DRYWELL) REQUIRED TO SURVIVE A ilYDROGEN BURN I)ist. from Qualification Equipment Center Line or Design Identification Function Elevation Azimuth of Reactor Temperature Duration Residual lleat Removal System (RllR) E12-F042A(1) LPCI-A Injection Valve (H.O.) 144'-3" 39 46'-0" 200*F 200 Hours EI2-F028A(1) Containment Spray
- Valve (H.O.) 170'-9" 30 59'-0" 200*F 200 Hours E12-F042B(1) LPCI-B Injection Valve (H.O.) 137'-10" 325 59'-0" 200*F 200 Hours E 12-F02811( l ) Containment Spray Valve (H.O.) 170'-9" 330 59'-0" 200*F 200 Hours ComI.astible Gas Control System (CCGS)
E61-C003A Recombiner 208'-10" 130 57'-0" 316*F 330 Days E61-C00311 Recombiner 208'-10" 330 57'-0" 316*F 330 Days E61-C00lA Purge Compressor 184'-6" 135 37'-0" 192*F(3) 22 Ilours E61-C00lB Purge Compressor 184'-6" 300 33'-0" 192*F(3) 22 Hours E61-F004A Swing Check Valve Vacuum Relief 194'-0" 220 33'-0" 350*F E61-F004B Swing Check Valve Vacuum Relici 194'-0" 220 33'-0" 350*F E61-t005A Ilutterfly Valve (H.O.) 194*-0" 240 33'-0" 200*F 200 llours E61-F005B 11ut ter f ly Valve (H.O. ) 194'-0" 240 33'-0" 200*F 200 llours E61-F001A Check Valve Vacuum Breaker 195'-1" 135 37'-0" 200*F 200 llours E61-F001B Check Valve Vacuum Breaker 199'-6" 300 33'-0" 200*F ' 200 llours E61-F002A Check Valve 195'-1" 135 37'-0" 200*F 200 llours E61-F002B Check Valve 198'-7" 298 30'-0" 200*F 200 Hours E61-F003A Butterfly Valve (H.O.) 195'-1" 135 33'-0" 200*F 200 Hours E61-F003B Butterfly Valve (H.O.) 198'-3" 298 28'-0" 200*F 200 Hours E61-D124 Ilydrogen Igniter 136'-0" 20 51'-9" 330*F 3 llours E61-D125 Ilydrogen Igniter .132'-11" 47 53'-0" 330*F 3 llours E61-D126 Ilydrogen Igniter 134'-4" 75 51'-9" 330*F 3 llours E61-D127 Ilydrogen Igniter 134'-4" 107 51'-9" 330*F 3 llours E61-D128 Ilydrogen Igniter 132'-10" 135 51'-9" 330*F 3 Hours
TABLE 5.6-4 (Continued) GRAND GULF NUCLEAR STATION CONTAINMENT EQUIPHENT (OUTSII)E DRYWEI.L) REQUIRED TO SURVIVE A IlYDROCEN BURN l Dist. from Qualification Equipment Center Line or Design Identification Function Elevation Azimuth of Reactor Temperature Duration l Combustible Gas Ceatrol System (CGCS) (Cent'd) E61-DI29 Hydrogen Igniter 132'-10" 165 51'-9" 330*F 3 Hours E61-D130 liydrogen Igniter 132'-10" 195 SI'-9" 330*F 3 Ifours E61-Dl31 Ilydrogen Igniter 145'-7" 220 60'-0" 330*F 3 liours E61-Dl32 Ilydrogen Igniter 134'-4" 253 51'-9" 330*F 3 Hours E61-D133 Ilydrogen Igniter 134'-4" 285 51'-9" 330*F 3 Ifours E61-Ill 34 Ilydrogen Igniter 134'-2" 317 52'-8" 330*F 3 llours E61-Dl35 Ilydrogen Igniter 136'-0" 349 51'-9" 330*F 3 llours E61-Dl36 Ilydrogen Igniter 166'-0" 16 SI'-9" 330*F 3 Hours E61-DI37 Ilydrogen Igniter 160'-4" 36 53'-6" 330*F 3 llours E61-D138 Ilydrogen Igniter 157'-10" 70 SI'-9" 330*F 3 Ifours E61-Dl39 Ilydrogen Igniter 157'-10" 100 $1'-9" 330*F 3 llours E61-Dl40 Ilydrogen Igniter 160'-4" 135 51'-2" 330*F 3 llours E61-Dl41 Ilydrogen Igniter 155'-10" 164 51'-9" 330*F 3 liours E61-D142 Ilydrogen Igniter 155'-10" 196 51'-9" 330*F 3 llours E61-Dl43 Ilydrogen Igniter 165'-0" 226 61'-4" 330*F 3 llours E61-Dl44 'ydrogen Igniter 160'-4" 260 54'-2" 330*F 3 llours E61-Dl45 lydrogen Igniter 159'-4" 285 51'-5" 330*F 3 llours E61-Dl46 Ilydrogen Igniter 159'-4" 321 51'-5" 330*F 3 Hours E61-Dl47 Ilydrogen Igniter 166'-0" 344 51'-9" 330*F 3 liours E61-0148 Ilydrogen Igniter 182'-10" 30 61'-0" 330*F 3 Ilours E61-DI49 Hydrogen Igniter 167'-8" 41 37'-0" 330*F 3 Hours E61-Dl50 Ilyd rogen Igniter 168'-10" 70 46'-2" 330*F 3 llours E61-l)l51 Hydrogen Igniter 168'-10" 109 51'-6" 330*F 3 llours E61-D152 Ilydrogen Igniter 178'-10" 70 46'-2" 330*F 3 llours E61-D153 Ilydrogen Igniter 178'-10" 109 51'-6" 330*F 3 llours E61-D154 Ilydrogen Igniter 182'-4" 136 51'-9" 330*F 3 flours E61-Dl55 Ilydrogen Igniter 182'-4" 254 55'-9" 330*F 3 llours E61-Dl56 Ilydrogen Igniter 183'-4" 278 47'-8" 330*F 3 llours E61-D157 Ilydrogen Igni ter 182'-4" 293 58'-11" 330*F 3 Hours E61-Dl58 Ilydrogen Igniter 183'-4" 320 53'-2" 330*F 3 llours E61-ill59 Hydrogen Igniter 202'-0" 21 50'-4" 330*F 3 Hours m m m m m m m m m m m m m
TABLE 5.6-4 (Continued) GRAND GULF NUCLEAR STATION CONTAINHENT EQUIPHENT (ODTSIDE DRYWEl.L) REQUlHED TO SURVIVE A IlYDROCEN BURN Dist. from Qualification Center Line or Design Equipment Duration Identification Function Elevation Azimuth of Reactor Temperature Combustible Gas Coxtrol System (CCGS) (Cent'd) 202'-0" 32 42'-0" 330*F 3 Hours E61-D160 Ilydrogen Igniter Ilydrogen Igniter 207'-9" 59 44'-2" 333*F 3 Hours E61-D161 3 Hours flydrogen Igniter 202'-0" 74 55'-8" 330*F E61-DI62 3 Hours Ilydrogen Igniter 202'-0" 88 48'-0" 330*F E61-D163 3 Hours Ilydrogen Igniter 202'-0" 92 48'-0" 330*F E61-D164 3 Hours Ilydrogen Igniter 202'-0" 106 55'-8" 330*F E61-Dl65 3 llours E61-D166 Ilydrogen Igniter 202'-0" 0 45'-0" 330*F 202'-0" o 37'-0" 330*F 3 llours E61-Dl67 Ilydrogen Igniter 330*F 3 flours Ilydrogen Igniter 202'-0" 0 34'-0" E61-Dl68 330*F 3 llours E61-D169 Ilydrogen Igniter 202'-0" O II'-0" Ilyd rogen Igniter 207'-8" 135 49'-10" 330*F 3 llours i E61-DI70 330*F 3 Hours E61-Dl71 Ilydrogen Igniter 208'-5" 210 49'-6" 204'-11" 242 26'-8" 330*F 3 Hours E61-Dl?2 Ilydrogen Igniter Hydrogen Igniter 204'-0" 256 53'-9" 330*F 3 llours E61-Ul?3 330*F 3 flours E61-Dl?4 Ilydrogen Igniter 204'-11" 284 53'-9" Hydrogen Igniter 204'-11" 298 26'-8" 330*F 3 llours E61-Dl75 330*F 3 flours i E61-D176 Ilyd rogen Igniter 207'-9" 310 56'-6" , 202'-0" 341 55'-1" 330*F 3 llours E61-Dl?7 Ilydrogen Igniter 330*F 3 llours E61-Dl78 Ilydrogen Igniter 262'-0" 5 55'-5" Ilydrogen I gn i t e.- 262'-0" 48 55'-5" 330*F 3 Ilours E61-Ul?9 330*F 3 Hours l E61-DI80 Ilydrogen IRuiter 262'-0" 91 55'-5" 262'-0" 140 55'-5" 330*F 3 Ilours E61-Dl81 Ilydrogen Igniter E61-D182 flydrogen Igniter 262'-0" 183 55'-5" 330*F 3 llours 262'-0" 225 55'-5" 330*F 3 llours E61-DI83 Ilydrogen Igniter 3 llours E61-Dl84 Ilydrogen Igniter 262'-0" 268 55'-5" 330*F 262'-0" 323 55'-5" 330*F 3 Hours E61-D185 Ilydrogen Igniter 3 Hours E61-pl86 Ilydrogen Igniter 283'-10" 349 39'-9" 330*F 283'-10" 34 39'-9" 330*F 3 Hours E61-Dl87 Ilydrogen Igniter 3 llours E61-Dl88 Ilydrogen Igniter 283'-10" 81 39'-9" 330*F 283'-10" 128 39'-9" 330*F 3 Hours E61-Dl89 Ilydrogen Igniter 330*F 3 Hours E61-Ul90 Ilydrogen Igniter 283'-10" 152 39'-9" 6 g M g g g g g g g g g
TABLE 5.6-4 (Continued) GRAND GULF NUCLEAR STATION CONTAINHF.NT EQUIPMENT (OUTSIDE DRYWELL) REQUIRED TO SURVIVE A HYDROGEN BURN Dist, from Qualification Equipment Center Line or Design Identification Function Elevation Azimuth of Reactor Temperature Duration Combustible Gas L Cectrol System (CGGS) (Cent'd) 3 Hours E61-Dl91 Hydrogen Iguiter 283'-10" 199 39'-9" 330*F E61-Dl92 Hydrogen Igniter 283'-10" 242 39'-9" 330*F 3 Hours E61-DI93 Hydrogen Igniter 283'-10" 286 39'-9" 330*F 3 Hours E61-Dl94 Ilydrogen Igniter 295'-0" 349 15'-3" 330*F 3 Ifours E61-Dl95 Hydrogen Igniter 295'-0" 159 15'-3" 330*F 3 Hours Transformers for Igniters Respective Locations 400*F(l) Containment and Dryvell Instrumen-l tation and Control System H71-TE-N007A Temperature Honitor 135'-4" 40 57'-0" 340*F 6 Hours H71-TE-N007B Temperature Honitor 135'-4" 205 57'-0" 340*F 6 Hours H71-TE-N007C Temperature Honitor 135'-4" 130 57'-0" 340*F . 6 Hours H71-TE-N007D Temperature Monitor 135'-4" 307 59'-0" 340*F 6 Hours Hil-TE-N009A Temperature Honitor 133'-0" 45 57'-0" 340*F 6 Hours H11-TE-N009B Temperature Honitor 133'-0" 214 57'-0" 340*F 6 Ilours H11-TE-N009C Temperature Honitor 133'-0" 125 57'-0" 340*F 6 Ilours H11-TE-N009D Temperature Honitor 13 3 ' - o 305 57'-0" 340*F 6 Ilours l Area Radiation Henitoring System D21-RE-N048B Area Radiation Honitoring 208'-10" 275 62'-0" 340*F 6 Hours D21-RE-N048C Area Radiation Honitoring 208'-10" 95 62'-0" 340*F 6 Hours Containment llatches
- c
- d Locks H23-Y002N Co Jainment Personnel Lock 212'-8" 140 62'-0" H23-Y00lN Conta isument Personnel Lock 124'-8" 130 62'-0" H23-Y003 Equipment Hatch 172'-3" 240 62'-0" M M M m M M M M m m W m m M'm m m m
TABLE 5.6-4 (Continued) GRAND GULF NUCLEAR STATIO?! CONTAINHENT EQUIPMENT (OUTSIDE DRYWELI.) _ REQUIRED TO SURVIVE A ilYDROGEN BURN Dist. from Qualification Equipment Center Line or Design Identification Function Elevation Azimuth of Reactor Temperature Duration Various Systems
; Level Transmitter (2) (2) (2) 350*F 10 Min.
Temp. Transmitter (2) (2) (2) Press. Transmitter (2) (2) (2) 318*F 26 Hin. Control Cables 346*F 3 Hours, 4 20 Hin. i Instrument Cables 340*F 6 Hours Power Cables 346*F 3 Hours, 20 Hin. Thermocouple Ext. Wire 340*F 5 Hours Terminal Blocks 340*F 5\ llours NOTES: A.O. = Air Operated H.O. = Hotor Operated i (1) Underwriters Laboratory approved maximum temperature for continuous operation at rate electrical load. (2) In various locations above t he 1100 floor. j (3) After 22 hours in a 192*F ambient, the steady state temperatures of various components are substantially below the maximum recommended temperatures. It is concluded that a 200*F ambient is still acceptable. l I i l l t m:Ca e e e g g g g g g g g g g g
.I Appendix A l lI
- I i
i,
- ,I
,!I !I l CONTAINMENT PRESSURE AND TEMPERATURE I I RESPONSE TO HYDROGEN COMBUSTION FOR CLEVELAND ELECTRIC ILLUMINATING CO. PERRY NUCLEAR POWER PLANT iI 1 I j OPS 38A92 i
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!! I .- !I CONTAINMENT PRESSURE AND TEMPERATURE II RESPONSE TO HYDROGEN COMBUSTION I FOR I CLEVELAND ELECTRIC ILLUMINATING lI PERRY NUCLEAR POWER PLANT
,I 'I OPS-38A92 ,
lI OCTOBER 7,1982
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I Offs $ ore PowerSystems , COVER SHEET DOCUMENT NUMSE R R EVISION DATED LEVEL & CONTROL ORDER NO. A487 OPS-38A92 A 10/7/82 C TITLE I CONTAINMENT PRESSURE AND TEMPERATURE RESPONSE TO HYDROGEN COMBUSTION FOR CLEVELAND ELECTRIC ILLUMINATING l PERRY NUCLEAR PLANT I I by l S. L. JONES G. M. FULS
.I 0FFSHORE POWER SYSTEMS P. O. BOX 8000 JACKSONVILLE, FL 32211 I
NUCLEAR SAFETY I RELATED I $4!N ATUR E/DATE P,:QJECT COORDINATOM M /*)l$1
&MANAGEfm /0/'la -l %~4%c% ' le' u AUTHORn '
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TABLE OF CINTENIS PAGE E INTRIUCTICN 1 mm 1 CNEi: DESCRIPTICN 1 INPUT INECRRTICN 2 l RESULTS 5 i SGt9Jcf 7 REFERE2CES 7
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1I - . u I ... LIsr OF TABLES i- : PAGE Tabla 1 MMQi Reactcr Coolant Mass and Energy Release Rates 8 Table 2 MMCH Hydrogen Release Rates and Tenperatures 9 Table 3 MMQi Fission Prtduct Energy Release Rates 10 Table 4 Barn Parameters 11
. Table 5 Canpartment Initial ccrditions 12 Table 6 Flow Path Parameters 13 r Table 7 Drywell Purge Systen Parameters 14 Table 8 Supiressicn Pool Parameters 15 Table 9 Spray System Parameters 16 Table 10 02npartment Dependent Passive Heat Sink Parameters 17 Table 11 Material Dependent Passive Heat Sink Parameters 18 Table 12 Drywell Passive Heat Sinks 19 Table 13 Wetwell Passive Heat Sinks 20 Table 14 Cbntainment Passive Heat Sinks 21 Table 15 Upper Pool Parameters 22 t
("I Table 16 Drawdown Parameters 23 i Table 17 Perry CI.ASIX-3 Results 25 r Table 18 C.ASIX-3 Results Ctmparison 26 LI "I - l i O - [. . . _ __ . , . . _ .
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II~ . p LIST OF FIGURES
!E
'E FIGURE Mark III Ci:mtaiment
'I Perry GASIX-3 Model 1
2 Perry SCRV Case: Drywell Tenperature 3
, Wetwen Tenperature 4 l Cbntalment Tenperature 5 Drywell Pressure 6 Wetwil Pressure 7 Contaiment Pressure 8 Ikywell Minus (bntalment Differential Pressure 9 Drywell 0 Gas Concentraticn 10 2
Wetu ll 0 2Gas Cbn::entration 11 e Contaiment 02Gas Concentraticn 12 Ikywell N Gas Cbncentraticn 13 2
- Wetwell N2 Gas Concentration 14 contalment N 2 Gas Crad.raticn 15 Drywell H2Gas concentraticn 16 Wetwell H2 Gas Cbncentration 17 Contalment H2Gas Concentraticn 18 Drywell Stem Gas Cbncentraticn 19 I Wetwell Steam Gas Concentraticn C%mtaiment Stean Gas Cbncentration 20 21 Perry DE Case
I Drywell Tenperature 22 a Wetwell Tanperature 23 [ C%mtaiment Tenperature 24 Drywell Pressure 25 g m eu m essure , Ccmtaimarit Pressure 27 Drywell MLnus Cbntalment Differential Pressure 28 Drywell 0 Gas Concentraticn 29 2 Wetwell ogGas Cbncentration 30 Ccntalment O2 Gas cancentraticn 31 I m
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% _ FIGURE DE Case (ccntinued):
Dcywen N2 Gas Cbnc=Rration
- 32 Wetwell N2 Gas Concentraticn 33 Otmtaiment Ny Gas Cbncentraticn 34 Drywell H2Gas Concentraticn 35 Wetwell H2 Gas Cbncentruticn 36 Centainnent H2 Gas Concentration 37 i
Dcywell Stean Gas Cbncentration 38 Wetwell Steam Gas Concentration 39 Cbntaiment Stean Gas Cbncentraticn 40
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INTROCXXTICN f At the regoest of Cleveland Electric nluninating, two analyses wre g performai f2 the Perry Nuclear Power Plant to investigate contalment E pressure and tamperature responses to degraded core even*J with hydrogen release and deflagraticm. The tw analyses differal in that cne was a stuck open relief valve loss of coolant accident (Iom) and the secx:nd cne us a drywen break Iom. Details of these tw cases are given later in this report. The CIASIX-3 program ue414=3 for these analyses is the same as that use$ for Mississippi Power and Light in supsort of their licensing activities for the Grand Gulf bbelear station. A diagram of the Mark HI Ctantalment and a schematic diagram of the Perzy CIASIX-3 nodel used in this analysis are given in Figures 1 and 2, respectively. There are three coupertments in this nodels the drywen, wetwen arxi ocntaiment. Alas includal is the suppression pcci, cx:ritaiment spray system, upper pool, and drywl1 purge system. The arrows in Figure 2 represent ficw paths tot a conpartments with the arrowhead pointirx3 in the directicn of allowed flow. CASE DESGIPTICN Two CIASIX-3 runs wre nude for the Perry Ibclear Plant. The ire.it for these the cases were identical except fer drawtw1 arri the locaticn of the steen, hydrogen and fissicr1 Ircxtuct energy releases. In the stuck cpen relief valve (SCRV) 10 3 case, the releases entered directly into the wetwen side of the suppressicri scol cuer the entire transient. In the i drywen break (DWB) IDG case the releases initiany entered only the drywan. At twenty minutes after initiation of the transient the steam, hydrogen ard fissicn pecduct energy releases were split, with half of the releases entaring the watwn side of the suEpressicn sool via the relief
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valves at:1 the other hal.f entaring the drywn. Although nors than half of j the releases are ex;=cted to discharga to the suppressicri pool through the depressurizaticn system, this 50/50 split was used as an estimata. Also at 1 1
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twenty minutes into the transient, the two C2nbustible Gas Cbntrol Systen (CGCS) cunpessaors were manually activated ard began punping gasses frcm the acr*=i=-st to' the drival.L After thirty minutes of transient, the upper pool began d= irs water to the suppressicn pool through cne line and
, cxmtinued dLrping for 8.67 minutes. 'Ihe drawdown of the suppressicn paol (reinstatenant of injection systems) was initiated at 6500 seo:rds into the transient. Releases in both cases '. ore continued until hydrogen equivalent to a 754 fuel clad metal-water reacticn was released fran the primary systam. At this time, the .TRV transient was terminated but the I55 transient was cmtinuei in order to allcw fcr a drywell burn.
I INPUT INFORMNTICN thless otherwise stated, Gilbert 02mmwealth suglied all of the input infonnaticn which is presented in this secticn. 'Ihe input paraneters' are specific to the Perry Plant but in nany cases are similar to those used in the CLASIX-3 analyses of the Grarri Gulf Nuclear Staticn. I Steen, hydrogen and fissicn grcxluct energy releases were taken fran a MAROI ccmputer code rtn provided by Battell; .%wial Institute of Chlum-bus. 'Ihe MAac21 results were nodified as discussed in Reference 1, Secticn III. These data were the same as that used in a similar analysis performed for the Mississippi Power and Light Grand Gulf 1bclear Staticm and are
. shown in Tables 1, 2, and 3.
1 \ j B.trn parameters are given in Table 4. 'Ihese control dien, Wtere and (- hcw much burnity occurs. Burns can be ignitad in any of the three ocmpart-ments and are allcwed to gropagate fran adjoining cranpartments through connecting ficw paths. These burn parameters are the hydreigen volune fracticn required for igniticn, the hydrcgen volune fracticn required for l propagaticn of a burn, the fracticn of hydrogen burned, the mininun crygen l g volune fracticn required for igniticn, the minimun oxygen volune fraction W required to swet ccnbusticn, the burn time ard the propagation delay time (dtich is ficw path depardent). 'Ihe burn parameters for this analysis were suggested by Offshore Pcwer Systans, agreed to by Gilbert Ccumcrwealth and are typical of grevious analyses. 7
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Parameters.fbr the cmpartment initial conditicms are given in Table
- 5. These includ's the net free gas volunes, the tenperatures, ard the oxygen, nil % and steam partial ;ressures. Partial Eressures wre calculated frcm ccupartment tanperatures, pressures and relative hunidities m=ing the contairment atmos;here consisted of a mixture of standard air
- stes..
I
':hore are tus flow paths incluied in this nodels watwell to contain-ment (WW-cr) and ocntalment to drywen (Cr-DW). 'Ihe ==v4== ficw area, ficw less coefficient, and burn Iropogaticn delay time for the W-CX2rr flow path are given in Table 6. 'Iha CP-IN ficw path consists of the drywen i purge systen. 'Ibe drywell purge system parameters are given in Table 7.
These include the sucticn cxmpartment, sink ocmpartment ard initiaticn time. 02ngressor head /ficw curves are given to allow for a variable flow rate depending cn the pressure differential between the contalment and
' drywell. 'Ibe contaiment vacuun relief valves are not nodeled in the. Perry analyses. Pricr analyses slow that the ocntainment pressure never goes l belcw atnespheric pressure, therefore the contaiment vacuan relief valves would not operata if they were nodeled.
Table 8 gives the suppressicn psol parameters, incluiing the initial iI pool water density, mass, tenperature and heat capacity. Gecmetry related scol peraneters are the number of vents, the flow area and largth of each vent, the subnergence depth to the bottan of the vent, turning loss j coefficients, gas loss coefficients and zdditicnal vent lengths to account fer fluid acceleraticn. 'Ihm pool surface areas in the drywen and wetwen L l are also included. 'Ihe weir height above the water level and the drywen holdup voltane ard surface area are necessary input parameters fcr the analysis of reverse flow through the su;pressicn pool. During reverse flow, water fran the suppressicn sool can overficw the weir wan arx1 renain g in the drywo u. L tI ;
'Ihe spray systau provides spray to the contairment with part of the s; ray ocetinuing through the wetwen. Scme of this wetwen s;rsy is in droplet form Wtile another fraction falls frca ledges as a sheet of wter.
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l j The remining fracticn of the contairment spray cannot enter the wetwen ii Ne== it m11ec.ts -; in the upper paol ard is drained directly into the suppressicn pool. A ratio of areas is used to cniculate the rates of ficw fcr the drain, droplet ard sheet. The spray ficw entering the wetwell as a sheet win be less effective than the droplet ficw but can be expected to have scrne cooling capability. It was assumed for this analysis that the sheet ficw is half as effective as the droplet ficw. Table 9 gives the irput parameters for the syray systen. The drcp diameter, spray tenpera-ture, and spray ficw rate for the cr.ntainment spray are specified. The drop size ard ficw rate exitire the contairment were used as the spray etnii-tiens for the wetwen. Cnly the fall tira and film coefficient are spe-cified fcr the wetwen spray. The fan times were based cn a tenairal velocity of 4.2 feet per secmd and the average spray fan height. Initia-I tien of the spray occurs after the first burn ard cxmtinues throughout the transient. Simulaticr1 of a drywen spray as used during the IHB case. The spray was initiated after an of the hydrogen was releasai and modeled safety injecticn ficw cut of the break. This cools the drywell atmosphere faster i (ard nere ecencmicany) than runnirg CCASIX-3 to cool the atmosphere by heat sinks cnly. This cooling condensid the steam and allowed a final hydrogen burn in the drywen. Table 9 s}ows the spcay paraneters. The passive heat sink data are given in Tables 10 to 14 inclusive. The I ccmpartment dependent heat sink parameters are fotrd in Table 10. Included are the initial heat sink tamperatures and radiant heat transfer beam lengths, which are based cn general geonetery censiderations and catain-ment dimensicrn. Table n gives the natarial dependent heat sink param-eters which are the anissivity, thennal carductivity, voltanetric heat c=p= city ard exit heat transfer coefficient. Tables 12,13 ard 14 give the passive heat sinks fcr the drywen, wetwen ani containnent respectively. The ntsnber of nodes per layer of passive hest sink is based cn the follcw-ing criteria. I 1. All coating layers have two rrdes. I
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- 2. All other layers have a minimun of three nodes with the actual I rumber be#g based en the thickness.
- 3. Steel milm have a spacing of apsroximately .02 inch per node for all thicknesses.
- 4. Ckmcrete walls have apeings of about one inch per node for the first six inches, two inches per node fcr the ne;tt twelve inches and six inches per node for the next one and a half fest. Beymd this, the wall is assumsi to be mai=htic.
Fbr ocmservatisu, the outer contaiment and wetwell all is assuned to be adiah tic after the steel layer. I 'Ihe upper pool and rehted parameters are given in Table 15. Rese include the locaticn of the upper pool, the volune dunped, the tanparature of the pool ater, the dunp flew rate and time of initiation of the dunp. The durp ficw rata is based cm an 8.67 minute durp time through cme line.
'Ihe dradown parameters are given in Table 16. Rese incitde the destinaticn of the fIcw, the voluna of water ratoved, ard the starting and ccrupletion times of the drawdom. Se drawdown for the DG case was simulatai by first filling the reacta vessel the fillirg the holdup volune. Se drawdown for the SORV case us simulated by filling the reactor vessel cnly.
RE:SULTS i i i A annary of the results of the two na **
- is given in Table 17.
Tenparature and pressure infonnaticn is given in Figures 3-9 fx the SCRV _ (stuck cpen relief valve) case and Figures 22-28 for the De (drywell breek) case. Plots of the volune fractions of cuygen, nitrogen, hydrogen, and staan are shown in Figures 10-21 for the !DRV case and Figures 29-40 for the De case. Table 18 gives the results of two similar analyses per-1 1- 5 j ?-
formed as part of the sensitivity sttz$y of the Grand Gulf Nuclear Station (see Referenew 1). '!he sensitivity ases in Reference 1 are considered to _ be generally applicable to BE Maz$c III cxmtairments. I '!he results of both the Perry and Grand GkAf SOW cases are similar. See Table 16 '!he contairment volme in the Perry Plant is analler than l Grand Gulf by 234. 'Ihis centributes to the extra contairment burn in the Perry transient. Ibwever, the Perry Plant has a 20% larger initial wetwen voltne than Grand Gislf 41ch results in fewer wetwsn hzens in the Perzy transient. Ibr the Perzy SOW case, peak taperatures and pressures cccurred in au conpart:nents during the first of the two contairment burns, at approx-imately 6900 seemds into the transient. '!he first a:ntairnent burn resultmi in the most severe pressure and tanparature excursion her anam wtwen igniticn occurred just before and during the contairnent burn. nb I wetwen igniticri occurred durirsg the second cxmtairnent burn due to a lack of cuygen. Referring to Figure 4, fbut additicnal wetwen tmperature peaks (at approximately 4445, 6555, 6965, and 7220 seccnds) stand out above the rest. Sgrags are not initiated tutil after the first watwn burn, dtich erplains , why the first wetwen tsuperatura peak is higher than those which imedi-ately fonow. 'Ihe other three "atzsve average" watwen tsmperature peaks occur because ignition takes place at increasai hydrcgen concentrations due to insufficient oxygen concentration when the hydrogen concentration reachai the 8 "/o setpoint. I
- t. Peak pressures and tanparatures for the Perry SOW case are cmparable in magnitude to those of the Grand Gulf SCRV ceae, except fcr the wetwen I peak temperature. 'Ihe Perry wtwH tsmperature is signi Fie=My higher due to the previcualy discussai coincident ecmbusticn in the wetwen and
.contaimato 911ch did rot occur in the Grand Gulf S0W case.
The results of the Perry DB case are also quite similme to the corre.wiing Grand Gulf mae, as shown in Table 18. Again, fewer wetwen I I _ _ . _ - 6
-~
_.. ,_ m . . .
l burns are evident fcr the Perry DWB case due to the larger wetwell volme, i The cnly other significant difference between the results for the tw plants relates to the containnent burn. Se Perry DWB case originally did not have a containnent burn associated with the final drywell and wetwell burn. The volume fracticn of hydrcgen in the containnent just price to the final burn us 0.065. See Figure 37. 'Ihe voltane fracticn of hydrogen required fcr igniticn is 0.00. 'Ib be conservative, it was decided to force a containnent burn at this scint to cbtain peak tenperatures and pressures. This reducal ocxicentraticn forced burn resulted in lower peak tenperatures and gressures for the Perry IMB case. The total anount of hydrogen burned in the Perry SOIN transient as I 2011 lbs. ard in the DWB transient was 2290 lbs. These values corresprd to 77.0% and 87.6%, respectively, of the total anount of hydrogen that as available. The similar Grand Gulf cases show the SCRV case burning 2332 lbs. and the DWB case burning 2243 lbs., Wdch are 89.3% and 85.8% of the total hydrogel releases, respectfttily. The difference between the per-centage of hydrogen burned in the Perry and Grand Gulf S3RV cases is due to the significantly larger nmber of wetwell burns in the Grand Gulf case. I sumute Ebr the two Perry cases analyzed, the peak calculated containnent i i pressure was approximately 21 psig, and brief duraticn tenperature peaks ,i ranged frcan 643 F in the drywell to 760 F in the contaiment to 1762 F in the wetwell. Ccmpariscn between the Perry and Grand Gulf analyses show the l SORV and IMB transients to be similar with the differences explained by I plant gec dry, the forced ocntainnent burn at a lower hydrcgen concen-traticn in the Perry INB case, and coincident crznbusticn in the wetwell and ocntainnent durirrJ the Perry SCRV case.
.il.
- 1. "CIASIX-3 containnent Pesponse sensitivity Analysis for the Missis-l sippi Ibwer and Light Grand Gulf Nuclear Station", A. D. Gunter and
! Dr. G.M. Ibis, Report 1Arnber OPS-37A15, December 1982.
i 7
.~ -n _ ,- - -
g- -
'JABIE 1
~ ~ ~
. Perry CIASIX-3 Input MARCH Reactor Coolant Mass and Energy Release Rates TPE Sequence E Time steam nelease mate Energy aelease aate
- .. E (secords) (Ibn/sec) (stu/sec)
- .g 0 220 260000.0
!: 5 602 183.33 219450.0 902 188.23 226316.67 1204 130.12 157050.82 1789 122.8 148670.43 1803 120.82 146396.67 2707 74.79 93053.33 j 2994 48.35 62419.85 1 3601 27.71 38470.73 I 3631 4201 30.51 4.72 42323.33 6501.99 4504 2.40 3203.05 I 4541 6.919 10793.33 4858 6.87 12699.55 [_ 5158 2.28 3556.80 5458 0.14 202.05 5758 1.08 2015.3 6058 0.10 153.21 6359 4.25 6601.6 7807.13 4.25 6601.6 7807.14 0 0 lg e 8 -- ----- _ . . . . . . .. ,..My=-..... "--
* * ' ~ ~ * -- "- '
-~
T -- _-____ T_ ~~'~?'- -- - w - _ T W~ ' * - --
lI .~ Perry CIASIX-3 Input 1 MARCH Bydrogen Release Rates and Tenperatures l l TPE Segjence
; Time Hydrogen Release Rate
'Dunpegature l- (seconds) (lbr/sec) ( F) 0 0 61.24
} 1803 0 (1.24 2707 1.225 x 104 525.36 2995 3.85 x 104 606.09 3295 6.00 x 10 d 694.34 3601 0.0071 784.66 3631 0.0089 788.80 3901 0.0479 880.29 4201 0.0486 753.07 4541 0.3186 1115.69
- 3 4858 1.0415 1693.75 I-5158 5458 0.4905 0.0691 1109.04 875.86 5758 1.0177 1702.01
[. 6058 0.0556 1039.08 6359 1.0415 1808.8 7807.13 1.0415 1808.8
- 7807.14 0 61.24 I
I 1 I I I 5
I. TABLE 3
- I . Perry CIASIX-3 Input MARCH Fission Droduct Energy Release Rates
- I TPE Sequence I Time (seconds)
Energy Release Rate (Btu /sec) I O O 3631 0 4541 246.47 4 5458 1097.76 6358 1530.3
- 6359 1530.7 7807.13 1530.7 7807.14 0 I .
I -I I I
.I I
I I E _ . _ , .10 -
,_R..
-- . ~~
~
f.
'"_~ . _ . . . . .
* 'M* a
__~.%'~ s-N. . _ . , __ w -- sw - - ~^
I. t !: TABIE 4
. . . , Perry CIASIX-3 Input L
Burn Parameters
- H /F for ignition 0.08 2
. 82 /F for gropagation 0.08 H2 fraction burned 0.85 Minimtn 02 /F for ignition 0.05 Minimum 0 2
/F to support embusion 0.0 Burn time (sec)** 6.45/2.26/11.25 I
I iI
'I
*If one number is present, parameters are the same in all empartments; l3E otherwise they are listed by drywell/wetwell/contaiment.
** Based cn flame speed of 6 ft/s.
iI
!I I
lI - lI
!.I I
- c. . .L
*~
11
..~ ___ J. - - ----~- - '~~~ ~-
Ig-TAB 2 5 i
.., Perry CLASIX-3 Input Cmpartment Initial Conditions Drywell Wetwell Containment r Voltne (ft3) 277,685 181,626 959,388 .I Temperature ( F) 134 90 90 I 02Pressure (psia) 2.83 3.01 3.01 q I N2 Pressure (psia) 10.63 11.34 11.34 I H2 O Pressure (psia) 1.24 .349 .349
!I I 'I I. I I I I I , I _ 1 -0.-; c a A :.,e;.:: : _ 12 -- -
- -- - .-- J
l r TABLE 6 l Perry CLASIX-3 Input Flcw Path Pararneters t i WW-CENT CINP-DW 2 1 Maxinun Ficw Area (ft ) 3187 See Table 7 I f Flcw Ioss Cbefficient 5.0 " " Burn Propagati:n Delay Tine (sec)* 1.0 a = 1 I I 'I I I lI s
~
- Based cm flame speed of 6 ft/sec.
I I - I I I u I -
~
a.-a . ..a .
- =. = - = = = - -
' TABIE. 7 Perry C.ASIT.-3 Inp2t l Drywell Purge System Parameters
- Suetial Osupartment Contairrnent Sink Canpartment Drywell
'l Initiatien Head Flow Rate **
(inches of gol (cm) ! 84.830 790 112 847 755 136.973 725 152.149 695 170.438 660 177.054 625 184.836 580 192.619 547 198.456 495 198.456 467 I 1 % .510 200.401 202.347 410 377 340 202.347 315 1I
~
205.460 287
- 1. .
l
- Manual initiaticn when hydrogen am= Hates in the drywell to a ocmcentraticn of 3.0 percent by voltzne or 20 minutes post ICCA, I whichever ocmas first.
I **This flow rate is for cne ccanIressor. Men both asnIressors are activatal, this table is multipliai by tm. 1
'I 4
- I i
..-. 14 -- __m_____.._. . .. .. ... . . . _ . -
1 I. J. TABLE 9
. Perry CASIX-3 Inp.rt Spray Systan Parameters t
Cont /Wetwell Drywell
, Flow rate (GPM) 5250 14077 Temperature ( F) 132 175 Drop diameter (microns) 370 6350 Fall time (seconds) 13.1/3.1 20 Heat transfer coefficient n (BIU/hR - Ft F) 20 10 contalment to wetwen Carry Over Fraction .4669 Initiaticn * **
*Initiaticn occurs after first turn.
**Initia'dcn occurs after hydrogen release steps.
(Drywen breax case cnly.) - I 1 l l _
.I
. . . . . . . . 16
._ c - 4 . .. _
- *- ~' ' - * ~ * .
- ~ . . - e s*
m ., . _, ~~ =
~ - y _ __ - , - - -
^
"% O-_________.N------- - - - - -
"~~ ~~ *~
TABIE 10
'~~y Perry CIASIX-3 Input I Cmpartment Deperdent Passive Heat Sink Parameters
'I Parameter Ccurpartment V61ue Temperature Drywell 130 F Pedestal 136 F
'~ Biological Shield Wall 170 F Wetwell 90 F Contairunent . 90 F Radiant Heat Transfer Drywell (vertical) Beam Length Platforms, grating,
'g weir mat 25.69 ft 5
Drywell slab 51.39 ft Drywell (horizontal) Weir Wall (inside) 11.36 ft lI l Weir Wan (outside), vent structure 1.44 ft Pedestal Wau s 13.06 ft Biological Shield Wall, I Drywen Wall 13.81 ft Wetwell (vertical) Platforms, grating 17.78 ft Wetwen (horizontal) Coltzans, vessel, I personnel lockweu , drywell all, vent structure 12.33 ft
,I I
17
~~~~~ ~
TABIE 10 (Cont'd)
~ ~ '
c Perry CIASIX-3 Input f Canpartment Deperdent Passive Heat Sink Parameters I l l I Parameter Compartment Value l Radiant Beat Transfer Contaiment (vertical) [5 Bean IAtngth (cont'd) Drywell Slab, platforms, grating, misc. 44.83 ft I Contairment (horizontal) Drywell wall, columns, personnel lock shielding 12.33 ft Vessel 62.42 ft I
~
I I LI ' I . I . I ! I .I
~ ~ - * * - - . . .
h , _,---7-. ~
--~n._~" _-EN_,..-,.
BBIE 11
'~
'3 Perry CLASIX-3 Input I Material Dependent Passive Beat Sink Parameters Parameter Material Value q anissivity Chemtree j; 0.8 Concrete 0.8 i '~
Stainless Steel 0.5 g Carbon Steel !i 0.8 Galvanized Steel 0.5 Coating 0.8
.I Thermal Corguctivity (Btu /hr-ft- F)
Chantree _ Concrete 1.965 0.8 ' Stainless Steel 9.4 g Carbon Steel
~g 26.0 Galvanized Steel 26.0 Coating 0.4 Volumetfig Heat Capacity Chemtree 64.5 (Btu /ft - F) Concrete 29.0 Stainless Steel 53.7 'l m Carbon Steel 53.9 Galvanized Steel 53.9 ExitHeat3r fer Coefficient Coating to Steel 4X10 (Btu-hr-ft - Cbating to Concrete 4X10 (or Chantree)
Steel to Concrete 10.0
} (or Chantree) 8 Concrete (or Chentree) to 10 I Concrete (or Chentree)
Steel to Steel Last Iayer Adiabatic Wall 10 0 8 I I I 1 I r g 19 _ _ _ _ _ _-_ _ _ . _ _ _ _ - - - - - - - - ^ ^-
TABIE 12
. . . , Perry CIASIX-3 Input
,i i
Drywell Passive Heat Sinks Surface 2 Layer Iayer Layer Description Area (Pr ) Ntaber Material Thickness (PT) Platform Structural 24,500 1 Coating 6.667X10[f Steel 2 Carbon Steel 3.125X10 Grating 45,800 1 Galvanized Steel 1.042X10
-2 Weir Mat 2,603 1 Coating 2.629X10-3
!R 2 Concrete Concrete 0.5 1.0 5 3 l 4 Concrete 1.5
~3 Weir Wall, Inside 3,198 1 Coating 2.629X10 2 Concrete 0.75 Weir Wall, Outside 1,225 1 Stainless Steel 2.083X10[2 2 Concrete 7.292X10 Pedestal Walls 3,410 1 Coating 2 Carbon Steel 6.667X10[2 8.333X10 3 Cherntree 0.5 Chantree 1.0 I 4 5 Chantree 1.417 Biological Shield Wall 9,665 1 Coating 6.667X10[f 2 Carbon Steel 8.333X10_1 L 3 Chemtree 9.167X10
-3 I Vent Structure 1,760 1 2
3 Stainless Steel Carbon Steel Concrete 8.333X10 7.50X10 0.5
-2 4 Concrete 1.917 Drywell Wall and 16,700 1 Coating 6.667X10)2 Drywell Slab 2 Carbon Steel 2.083X10
.I 3 4
Concrete Concrete 0.5 1.877 I I II 2a
.I .. . . . =- -= :- . ._ _
I .
- 13
-g
-, . Perry CIASIX-3 Input
.I Wetwell Passive Heat Sinks Surface Layer Iayer Layer Description Area (M)_ Number Material 'Ihickness (PI)
Steel Coltans, 56,092 1 Coating 6.667X10j Platform Strt:ctural 2 Carbon Steel 5.199X10 Steel, Contairment Vessel Steel Coltans, 2,074 1 Stainless Steel 2.083X10
-2 lower 20' 2 Carbon Steel 9.167X10
-2 Grating 37,230 1 Galvanized Steel 1.042X10-2 Centairment vessel, 1,885 1 Stainless Steel 1.25X10
-1 lower 5' 2 Carbon Steel 1.125X10~1 Personnel Iock Well, 5,498 1 Ccating 7.125X10
-3
!, Drywell Wall 2 Concrete 0.5 3 Concrete 2.0 Personnel Iock Well ~4 753 1 Coating 6.667X10 2 Carbon Steel 8.333X10
-2
,. g 3 Concrete 0.5 E 4 Concret. 2.0 , Vent Structure 1,540 1 Stainless Steel 8.333X10
-3 2 Carbon Steel 7.50X10
-2
]' 3 Concrete 0.5 4 Concrete 1.917 fl. I O I I I #
- = = _ . = na :- m -
O 0I Perry CIASIX-3 Input Contairunent Passive Heat Sinks ~ Surface 2 Iayer Iayer Layer i Description Area (Pr_)_ Number Material thickness (Pr) E- Drywell Wall, 37,891 1 Coating 2.780X10
-3 !
l Drywell Slab, 2 Concrete 0.5 Fuel Transfer 3 Concrete 1.0 l f Floor, etc. 4 Concrete 1.259 l 4 3 Steel Columns, 11,015 1 Coating 6.667X10
-2
~5 Chk'd Plate 2 Carbon Steel 6.25X10 Platform Structural 38,275 1 Coating 6.667X10 4 Steel 2 Carbon Steel 2.292X10
-2 Grating 21,830 1 Galvanized Steel 1.042X10-2 Contairunent vessel 66,787 ~4 1 Coating 6.667X10 2 Carbon Steel 1.250X10-1
~
Fuel Transfer Floor 370 1 Stainless Steel 2.083X10
-2 l Slab 2 Concrete 0.5 3 Concrete 1.0 l 4 Concrete 1.5 3
Personnel Icck 610 1 Coating Shielding 2 Carbon Steel 6.667X10_f 2.083X10 [ 3 Chemtree 0.5 4 Chemtree 1.0 5 Chemtree 1.5 g I bI. e 2 c:. onn
TABIE 15 I . .. , .
= -
Perry CIASIX-3 Input
.W
~
Upper Pool Parameters j Iocation Containment
.. Voltane Dtroped (ft3) 32,830 Temperature (OF) 100 Ctzup Flow Rate (ft / min) 3,787 Initiation
- I
;I !I g
I
!'.I l-I
- Initiation occurs at 30 minutes after IIXA signal.
I
- I .
I I
,,,-,_.w,- -,m
l . t [I
- sr. TABIE 16 l i' I Perry CIASIX-3 Input l l
~
Drawdown Parameters 4 Destinatico of Reactor Drawdown Flow Vessel Drywell* , Volume Restwed (ft ) 13,939 40,564
,_ Starting Time (sec) 6,500 6,952.5
(. Ccapletion Time (sec) 6,952.5 8,269.5
- s I
g --- e me em u . _ m e . ca . M I . I I g 24 y~ . . . . . . - -.-
I. 1 TABIE 17 Perry CIASIX-3 Results SORV DWB Nunber of burns m* 0 0 [1] l W 32 30 [8] . C 2 0 [1]
.I Total H2Burned (lbn) W 0 0 [117] . W 1220 1361 [472]
CT 791 0 [340] l H2 Remaining (lbn) W 15 692 [203] W 293- 151 [ 41] C 294 409 [ 81] i Peak Tunp. (UF) N 191 ( 154) 251 [ 643] : W 1762 (1364) 1201 [1763] C 760 ( 236) 192 [ 587] Peak Pressure (psig) N 15.9 (10.7) 13.8 [17.3] W 21.1 (12.6) 12.2 [19.4] C 21.2 ( 9.9) 10.9 (19.4] l m
*Drywell, wetwell, and contaiment are abbreviated as W, W, and C.
( ) Maxima due to wetwell burns. [ ] Values due to extension past end of hydrogen release. I
~I I 3
_ _ _ : b __ . . _ _ _ _ ______ -
$. ., c ;
p . M M
~
M M M M A TABLE 18 , CIASIX-3 Results Camparison Perry Results Grand Gulf Results** i SORV DMI SORV DWB
. s Number of burne DW* O 4 [1] O O [1]
w 32 L [8] 59 26 [6] ,-' Cr 2 0 [1] 1 0 [1] Total H2Burned (lba) DN O O [117] O O [104] w 1220 1361 [472] 1820 1233 [319] Cr 791 0 [340] 512 0 [567] 11 , H2Remaining (1hn) DW 15 692 [203] 25 712 [240]
- l e w 293 151 [ 41] 40 21 [ 15]
- l. Cr 294 409 [ 81] 207 629 [114]
( Peak Temp. ( F) DW 191 ( 154) 251 [ 643] 193 ( 137) 2 % [ 707] W 1762 (1364) 1201 [1763] 1020 (1020) 1110 [2295] Cr 760 ( 236) 192 [ 587] 681 ( 197) 1 % [ 860] l! !- Peak Pressure (peig) DW 15.9 (10.7) 13.8 [17.3] 18.9 (9.6) 12.3 [16.3]
- w , 21.1 (12.6) 12.2 [19.4] 23.5 (9.0) 11.9 [31.6]
Cr 21.2 ( 9.9) 10.9 [19.4] 23.9 f,8.8) 11.7 [32.1)
*Drywell, wetwell, and contairunent are abbreviated as DW, W, and Cf.
( )Maxbna due to wetwell burns. [ ] Values due to extension past end of hydrogen release. ,
**See Reference 1, cases SA1 and DA4.
I, . ii l l f ..
- ' L J
! .1 J CONTAINMENT C . j
.- 4 E
~
, l UPPER POOL
. ? .-
.. J ., :
< :. . , . e.
S
., , _ r,
- r. . . .
I WETWELL l N % DRYWELL g p ,
~
s I. i , SUPPRESSION r C 2 POOL ! ....'Y.
.. - . E. . MARK lil CONTAINMENT t
b FIGURE 1 a E ..
.. a .. ~_ = -
-- - _ _ _:_ _ w . - . . ; ;=__. .
PERRY CLASIX-3 MODEL I
~
sPary canny ovan ?" _'J_'__'__ ;.
- WETWELL CONTAINMENT
-4 >
(Vol. 2) (Vol. 3) d O i t S r= ~I i m I UPPER PQQL Y DUMP : _ -3 l i EE:Eiiiiiiiiiililih
@Z:FeatsIL48M i
DRY WELL
----+i:=hisisisisisL<
~
--_} (( (Vol.1)
.;;_- - _ -;_ :; 1 Z
'-l hE[Z[Z{555'
> FLOW IN ONE DIRECTION
* * * * * * * *
- SPRAY HEADERS I FIGURE 2 I
i P 200.0,
.I >
c : t ' i : 180.0 : i c-e w - E : ' 52 s [E < e
'3 g 160.0 : - - -
i Z : g .
% D
- I 140.0' i
[ llE I1 lEl llE < -- r
-l i
L I : 0.0 2.0 4.0 6.0 8.0 TIME (SECONDS) X 1000 Lg CLEVELAND ELECTRIC ILLUMINATING
' PERRY NUCLEAR STATION SORY ORYWELL TEMPERATURE
.I FIGURE 3
. - +
. . . . . g .- ;; = = . -
)
~ ~ --
- ~ 4,- g - er r ---x.
- ' = =
r.
=
,L* L ~ '
, . , - ;-~
' "- ^'~:~_
*~-_ l' _'_ ' ~ L7_^
[g 2000.0 p.. , .
- f. C N
=
h 1000.0 E z - n .
~ >
g . D
, y ll tW (Ll bl.LI Q %*
0.0 0.0 2.0 4.0 g 6.0 8.0 TIME (SECONDS) X 1000 LE CLEVELAND ELECTRIC ILLUMINATING jg PERRY NUCLEAR STATION SORY WETWELL TEMPERATURE
;E lI I
FIGURE 4 I I [ -i -----:-.,,*.--g., a--. . *; 2 7
- T ' :.. _= * -
m 9
.I .I - ..
d
.l 800.0, 600.0 r
c LaJ K ' 400.0 ! I .h CL. z I 0.0 - l 0.0 2.0 4.0 TIME (SECONOS) X 1000 6.0 Y.0 l1-l CLEVELAND ELECTRIC ILLUMINATING l PERRY NUCLEAR STATION SORV CONTAINMENT TEMPERATURE ,l 1 lI FIGURE 5 I L.I A N~~ ~ ow *,e:t 'r*=,,Y$.,*."*~~ 7 ,7_ = 7_[ ~** *
!I '
il - . 40.0 , . Il t - l ,
-g q 30.0 ,
E ~ m . e . m . - E 0 ' Y
~ .
E 20.0 a
/
- -g '
/ .
5 :
/
/
10.0 '
,I 0.0 2.0 4.0 6.0 8.0 TIME (SECONDS) X ,1000 l
- 1. CLEVELAND ELECTRIC ILI.UMINATING PERRY NUCLEAR STATION SORV l DRYWELL PRESSURE I
~
l . FIGURE 6 s lI I
.I
~
L _ _ _____-__ n -e_-2_____ - - - -
lI t -I 40.0 .
- I F
- :
P , l l-l . l I
- 30.0 fB 5
- 1
, E
~
i . . L I 8 !l , 0 E 20.0 (:
-I :
1
@lg a jbb
(
-l I
10.0 0.0 2.0 4.0 6.0 8.0 j TIME (SECONDS) X 1000 )- l CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION SORY
,I VETWELL PRESSURE l.
Ll FIGURE 7 I E
,I il m.__._..__ _ _ _ - - -
3.. II . . . 40.0 7 1 i l
.g- i
,g ; 30.0 ,
W Di - s - l a E c - l E 20.0 i (.
\
h ' - l f 10.0- . l 0.0 2.0 4.0 6.0 8.0 TIME (SECONOS) X 1000
- L!
I CLEVELAND ELECTRIC ILLUMINATING' PERRY NUCLEAR STATION SORY CONTAINMENT PRESSURE
~
.I FIGURE 8
.-I ll
'I .
L . . . .
A 8*4 Se O 4 , gee gg ,
- y. .
lI . e
/ t g ,/ y ,d'y 'N 8
1 .
/ rr .
rI E 5 00
/ /~
E - I d .
'g y - ~l I -10.0- - ~
TIME (SE b DS) X 1000 g CLEVELAND ELECTRIC ILLUMINATING t_ . PERRY NUCLEAR STATION SORV g DIFTERENTIAL PRESSURE iL I FIGURE 9
-I I
=4e y _- --
- __ . - - W .- - '_____________--_-___
- - - - - - - - - - - - - - - - - - - - - - - ~
4 II . .., . ,. m
- 8 0.20-l .5 , .
1 'g : e y I:= "
/
,pg , - in E P 0.19 '
\
15 W . g . . I D D E , 3 .
.l g 0.18 ,
D l D 8 D .
'I 0. 17 - -
0.0 2.0 4.0 'l 6.0 8.0 TIME (SECONDS) X 1000 l.. l CLEVELAND ELECTRIC ILLUMINATING ,.g PERRY NUCLEAR STATION SORY DRYWELL 02 GAS CONCENTRATION iI &I iI -
. FIGURE 10
'I l.. I - 3:- _ . ,...- = . . --- .-- --- -
^ 2 .L _ -.. - - - -
.- -.m.mo se o e o e- .m. .-m
.....~..o*.v.,"ZJ. . .
, e O 0.30
.E 0.20 '
, -.I . m h
- 5 '
E .
- s. I .
\
E q g 0.10 lb . f ( . e - [- I : r6 O.00 I- - O.0 2.0 4.0 6.0 8.0 TIME (SECONOS) X 1000 '.I t l l I . CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION SORV l eI WETWELL G2 CAS CONCENTRATION l-e I
\
FIGURE 11
~ .I I 4
da O E' eMN ew4SS -- _g,g ,g, g
- g. .
0.30 . 1
;I ,
.. 0.20 ,
~
u I ! w .
-I s :
y 0.10 I . l.. I . a I 0.00 -- - ' TIME (SECb OS) X 1000 I CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION SORV CONTAINMENT 02 GAS CONCENTRATION I I FIGURE 12 I
.I
~I ______.. _ __ ____ ____ .____ _ e a . . , - .
E -
.l "l_
0.74; -
-- i
' 1
= ,
'. l z
o C [ - E / O.73 '
\ '
y ,' . ~ a I S l . I : 0.72 g 0.0 2.0 4.0 6.0 8.0 TIME (SECONDS) X 1000 l CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION SORY DRYWELL N2 I GAS CONCENTRATION 1 I I FIGURE 13 I II g - __, ,, _ -__* _U . v. ^.. . ..***u .. . . . er-
e -M&- 66 '.M66 Oge ea4 ee e . G em e gg lI ~
~
~
g 0. 0 . I 1 i : LI 5 e f 0.60 i ' ' f . r I w , o-l : f N \:. 0.40 l 0.0 2.0 4.0 6.0 8.0 TIME (SECONOS) X 1000 Ll CLEVELAND ELECTRIC ILLUMINATING
. PERRY NUCLEAR STATION SORY WETWELL N2 CAS CONCENTRATION I ~
FIGURE 14
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0
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E 0.70 . I m u
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h
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. f-I .
I D 0.Sb ' l 0.0 2.0 4.0 6.0 8.0 TIME (SECONOS) X 1000
- 4 CLEVELANO ELECTRIC ILLUMINATING PERRY NUCLEAR STATION SORY CONTAINMENT N2 GAS CONCENTRATION I
FIGURE 15 I I I
.a..- *
,m-_ _.' - .- :---
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i
/ -
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< 0.006 -
ll , w : F I E iE! 0.004 ,'
-l 0.002 g i :
x ;
/ .
1 0.000 l 0.0 2.0 4.0 6.0 8.0 TIME (SECONDS) X 1000 l CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION SORY DRYWELL H2 l GAS CONCENTRATION I I FIGURE 16 I
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I . FIGURE 17 I
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/ q l
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-l FIGURE 18 I
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= ~ . . - . . - -
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- 0.06 '
0.0 2.0 4.0 6.0 8.0 TIME (SECONOS) X 1000 I CLEVELAND ELECTRIC ILLUMINATING I DRYWELL PERRY NUCLEAR STATION SORY STEAM GAS CONCENTRATION I FIGURE 19 _I
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'"l WETWELL STEAM GAS CONCENTRATION g FIGURE 20 I
I * - * " -
e I' HI I r 0.50,. -
- g. .
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- i O.10 ,
~
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, 100.0' , '
O.0 10.0 l TIME (SECONOS) X 1000 20.0 l I. CLEVELAND ELECTRIC ILLUMINATING 9 l PERRY NUCLEAR STATION DWB ~ DRYWELL TEMPERATURE I il FIGURE 22 l l 1... l l l . Q * @O e U* * (= em em -- _._-._, * * *Wn* 1 "'*
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@ 1000.0 g .
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Q - L L t J5 0.0 0.0 10.0 20.0 TIME (SECONDS) X 1000 I CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION DWB WETWELL TEMPERATURE I 9 I FIGURE 23 I I I I .. . ..
h
~I " **
600.0, '
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20.0 TIME (SECONDS) X 1000 j_- CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION DWB g CONTAINMENT TEMPERATUAE I I FIGURE 24 I - I . I-
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l D L 10.0 - 0.0 10.0 20.0 TIME (SECONDS) X 1000 CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION DW8 DRYWELL PRESSURE I W I FIGURE 25 I I .
.- .e. e - me . ... e. . .
~
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p 30.0 E-g . 8 C : \ ; ( E 20.0 \ r- % %% I '( - _ _
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10.0 0.0 10.0 20.0 1 TIME (SECONDS) X 1000 CLEVELAND ELECTRIC ILLUMINATING g PERRY NUCLEAR STATION DVB WETVELL PRESSURE FIGURE 26 4 e E -
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, FIGURE 27 i
v - 9 9% Q . . -.2-__.... . . _ _. .- - - - - . , , ,_ ,-..,,,,,, , ,._ ,. .- ,-. --,. - _ . ,--.....- - _ -,---.-.---.,. - , - - - - . - - ,
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0.0 10.0 fl t. 20.0 TIME (SECONDS) X 1000 1I CLEVELAND ELECTRIC ILLUMINATING L.g PERRY NUCLEAR STATI0'N DWB DIFTERENTIAL PRESSURE I FIGURE 28 6 m
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. TIME (SECONDS) X 1000
'~ CLEVELAND ELECTRIC ILLUMINATING .l PERRY NUCLEAR STATION DWB DRYWELL 02 CAS CONCENTRATION I i FIGURE 29 R h" I I *
~
l
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~
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,I CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION DWB WETWELL 02 CAS CONCENTRATION I -
FIGURE 30 I I E, . I .-
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. 9
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g . l w 5 ~ E 0.10 .
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I 0.00 0.0 10.0 l 20.0 TIME (SECONOS) X 1000 l CLEVELAND ELECTRIC ILLUNINATING PERRY NUCLEAR STATION DWB I . CONTAINMENT 02 GAS CONCENTRATION I FIGURE 31 I I O *
.*Oa e
i~ I ~ ~ ' j'l 0.80. I
. . 0.60 !
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0.20 ! I 0.00 - 0.0 10.0 20.0 g TIME (SECONDS) X 1000 'l- CLEVELAND ELECTRIC ILLUMINATING. PERRY NUCLEAR STATION DWB g DRYWELL N2 GAS CONCENTRATION I
;'I rieaae 32 t._ , .I .
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#+ + IMAGE EVALUATION ,f8g
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- f I 20.0
, TIME (SECONDS) X 1000 CLEVELAND ELECTRIC ILLUMINATING g PERRY NUCLEAR STATION DWB i DRYWELL H2 GAS CONCENTRATION l 1 hI FIGURE 35 I l, I
.-9 s ---_5--&--4 &--" -
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. FIGURE 36 s I
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0.02 : rl . t 0.00 0.0 10.0 20.0 TIME (SECONDS) X 1000 J[ 4l CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION OWB I CONTAINMENT H2 GAS CONCENTRATION i
. FIGURE 37 s
I lm g i
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me e 4 pe .e e 'mmM
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0.80 \ - p N 3 5 0.s0 x : a \ : I _ I 0.40 ' N : b k 4 N 0.20 l . ; 0.00' O.o 10.0 20.0 g TIME (SECONDS) X 1000 CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION OWB g DRYWELL STEAM GAS CONCENTRATION I FIGURE 38
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=
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9 *
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h [\ % ;
^
0.0 10.0 20.0 .l TIME (SECONDS) X 1000 CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION DWB l WETWELL STEAM GAS CONCENTRATION I I FIGURE 39 I I I *
.. = : - _ _ - . .
we . e ee e -,w.- e e o i + #. m - e e= m'sw- e N o e -e,- =
- me m em e ,
k l ' !I
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,I :
~
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0.00' i 0.0 10.0 20.0 TIME (SECONDS) X 1000 LI l CONTAINMENT CLEVELAND ELECTRIC ILLUMINATING PERRY NUCLEAR STATION DWB STEAM GAS CONCENTRATION I I I FIGURE 40
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