ML20237C074
| ML20237C074 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 08/31/1998 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20237C072 | List: |
| References | |
| NUDOCS 9808200188 | |
| Download: ML20237C074 (54) | |
Text
- ..
O Simplified Passive Advanced Light !
Water i Reactor Plant Program l
AP600 TIER 1 MATERIAL i
1 0
Prepared for l U.S. Department of Energy
! San Francisco Operations Office
- DE-AC03-90SF18495 i Revision 7 i August 1998 O
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9 Tier 1 Material TABL.E OF CONTENTS E 1 r^g Revision: 7 C Effecthe: August 1998 !
TABLE OF CONTENTS (cont.)
Section Page 2.6.3 Class IE de and Uninterruptible Power Supply System 2.6.3 1 2.6.4 Onsite Standby Power System 2.6.4-1 2.6.5 Lighting System 2.6.5-1 1 2.6.6 Grounding and Lightning Protection System 2.6.6-1 2.6.7 Special Process Heat Tracine System 2.6.7-1 2.6.8 Cathodic Protection System 2.6.8-1 2.6.9 Plant Security System 2.6.9-1 2.6.10 Closed Circuit TV System 2.6.10-1 2.6.11 Main Generation System 2.6.11-1 2.6.12 Excitation and Voltace Reculation System 2.6.12-1 2.7 HVAC Systems 2.7.1 Nuclear Island Nonradioactive Ventilation System 2.7.1-1 2.7.2 Central Chilled Water System 2.7.2-1 2.7.3 Annex / Auxiliary Building Nonradioactive Ventilation System 2.7.3 1
.2.7.4 Diesel Generator Building Ventilation System 2.7.4-1 2.7.5 Radiologically Controlled Area Ventilation System 2.7.5-1 2.7.6 Containment Air Filtration System 2.7.6-1 O
V 2.7.7 Containment Recirculation Cooling System 2.7.7-1 2.7.8 Radwaste Building HVAC System 2.7.8-1 )
2.7.9 Turbine Island Buildine Ventilation System 2.7.9-1 2.7.10 Health Physics and Hot Machine Shop HVAC System 2.7.10-1 2.7.11 Hot Water Heatine System 2.7.11-1 3.0 NON-SYSTEM BASED DESIGN DESCRIPTIONS AND ITAAC 3.1 Emergency Response Jacilities 3.1 -1 ;
3.2 Human Factors Engineering 3.2-1 3.3 Buildings 3.3-1 3.4 Initial Test Program 3.4-1 3.5 Radiation Monitoring 3.5-1 3.6 Reactor Coolant Pressure Boundary Leak Detection 3.6-1 3.7 Design Reliability Assurance Program 3.7-1 4.0 INTERFACE REQUIREMENTS 4.0-1 5.0 SITE PARAMETERS 5.0-1
- Underlined sections - title only, no entry for Design Certification.
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Tier 1 Mitirlil l
PASSIVE CONTAINMENT COOLING SYSTEM d $$
l Revision: 6
!j '
Effective: July 1998 s
l i L Table 2.2.2-3 (cont.)
Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria 6.a) The Class IE equipment Type tests or a combination of A report exists and concludes that identified in Table 2.2.2-1 as type tests and analyses will be the Class IE equipment identified being qualified for a harsh performed on Class IE equipment in Tables 2.2.21 as being environr.ient can withstand the located in a harsh environment. qualified for a harsh environment environmental conditions that can withstand the environmental i would exist before, during, and conditions that would exist before,
- following a design basis accident during, and following a design ,
l without loss of safety function basis accident without loss of l for the time required to perform safety function for the time the safety function. required to perform the safety function.
6.b) The Class IE components Testing will be performed by A simulated test signal exists at identified in Table 2.2.2-1 are providing a simulated test signal the Class IE equipment identified powered from their respective in each Class IE division. in Table 2.2.2-1 when the l Class lE division. assigned Class lE division is provided the teu signal.
l 6.c) Separation is provided See Tie. I Material, Section 3.3, See Tier 1 Material, Section 3.3, i between PCS Class IE divisions, Nuclear Island Buildings. Nuclear Island Buildings, and between Class IE divisions 1
and non-Class IE cable. '
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2.2.2-9 l
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Tier 1 M;t:ri;l PASSIVE CONTAINMENT COOLING SYSTEM ~
d "1 I
Revision: 7 Effective: August 1998 _
Table 2.2.2-3 (cont.)
Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment Inspections, Tests, Analvses Acceptance Criteria 7.a) The PCS provides the i) Testing will be performed to i) When tested separately, each delivery of water to the outside measure the PCCWST delivery of the two flow paths delivers of the containment vessel. rate from each of the two parallel greater than or equal to:
flow paths.
- 442 gpm at a PCCWST water level of 23.70 ft 2 0.25 ft above the lowest standpipe 123.5 Fpm at a PCCWST water level of 20.65 ft 2 0.25 ft above the lowest standpipe
- 72.5 Fpm at a PCCWST water level of 13.05 ft 2 0.25 ft above the lowest standpipe.
ii) Testing and or analysis will be performed to demonstrate the ii) When tested and/or analyzed PCWST inventory provides with both flow paths delivering 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of cooling. and an initial water level at 24.25 + 0.25, - 0.00 ft, the water inventory provides greater than or equal to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of flow with a flow rate greater than or equal to j 62.7 gpm.
iii) Inspection will be performed to determine the PCCWST iii) The elevations of the standpipes elevations. standpipes above the bottom standpipe are:
1 6.1 ft 0.25 ft 14.0 ft 0.25 ft
- 21.6 ft 0.25 ft O
2.2.2-10
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l Tier 1 Mit: rill PASSIVE CONTAINMENT COOLING SYSTEM M !!
Revision: 6 E Effective: July 191#8 l Table 2.2.2-3 (cont.)
Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria l 7.b) The PCS provides wetting i) Testing will be performed to i) A report exists and concludes I of the outside surface of the measure the wetted surface of the that with water in the PCCWST I containment vessel and the inside containment vessel from either of at the following levels, water I and the outside of the the two parallel flow paths to the delivery to the containment shell l containment vessel above the containment vessel. provides coverage measured at the I operating deck is coated with an spring line that is equal to or I inorganic zine material. greater than the corresponding coverage used to calculate peak containment pressure in the safety analysis.
l - 23.70 2 0.25 ft above the lowest standpipe
- 20.65 t 0.25 ft above the lowest standpipe I - 13.05 t 0.25 ft above the lowest standpipe ii) Inspection of the containment ii) A report exists and concludes I vessel exterior coating will be that the containment vessel conducted. exterior surface is coated with an I inorganic zine coating above l elevation 135'-3" l iii) Inspection of the containment iii) A report exists and concludes i vessel interior coating will be thet the containment vessel I conducted. interior surface is coated with an I inorganic zine coating above 7' I above the operating deck.
7.c) The PCS provides air flow Inspections of the air flow path Flow paths exist at each of the over the outside of the segments will be performed. following locations:
I containment vessel by a natural circulation air flow path from the - Air inlets air inlets to the discharge - Base of the outer annulus structure. - Base of the inner annulus
- Discharge structure 7.d) The PCS provides drainage Testing will be performed to With a water level within the of the excess water from the verify the upper annulus drain upper annulus 10" 1" above the outside of the containment vessel flow performance, annulus drain inlet, the flow rate through the two upper annulus through each drain is greater than drains. or equal to 450 rp.m.
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- _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ ____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _-_a
Tiir 1 Mit;rirl PASSIVE CONTAINMENT COOLING SYSTEM .iif
" 9E Revision: 7 Effective: August 1998 -
Table 2.2.2-3 (cont.)
Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment Inspections, Tests, Analyses Acceptance Criteria 7.e) The PCS provides a flow i) See item I in this table. i) See item I in this table.
path for long-term makeup to the PCCWST. ii) Testing will be performed to ii) With a water supply measure the delivery rate from the connected to the PCS long-term long-term makeup connection to makeup connection, each PCS the PCCWST. recirculation pump delivers I greater than or equal to 62.7 gpm when tested separately.
7.f) The PCS provides for long- i) Testing will be performed to i) With the PCCWST water level term makeup from the PCCWST measure the delivery rate from the at 23.75 ft 0.5 ft above the to the spent fuel pool. PCCWST to the spent fuel pool. bottom of the tank, the flow path from the PCCWST to the spent fuel pool delivers greater than or equal to 50 gpm.
ii) Inspection of the PCCWST ii) The volume of the PCCWST will be performed. is greater than 400,000 gallons.
8.a) The PCS provides a Inspection of the PCCAWST will The volume of the PCCAWST is PCCAWST initial inventory of be performed. greater than 363,000 gallons.
cooling water for PCS delivery from hour 72 through day 7.
8.b) The PCS provides the Testing will be performed to With PCCASWST aligned to the delivery of water frorn the measure the delivery rate from the suction of the recirculation PCCAWST to the PCCWST. PCCAWST to the PCCWST. pumps, each pump delivers l greater than or equal to 62.7 gpm when tested separately.
8.c) The PCS provides water See Tier 1 Material, subsection See Tier 1 Material, subsection l inventory for the fire protection 2.3.4, Fire Protection System. 2.3.4, Fire Protection System.
system.
- 9. Safety-related displays inspection will be performed for Safety-related displays identified identified in Table 2.2.2-1 can be retrievability of the safety-related in Table 2.2.2-1 can be retrieved retrieved in the MCR. displays in the MCR. in the MCR.
10.a) Controls exist in the MCR Stroke testing will be performcd Controls in the MCR operate to to cause the remotely operated on the remotely operated valves cause remotely operated valves valves identified in Table 2.2.2-1 identified in Table 2.2.2-1 using identified in Table 2.2.2-1 to to perform active functions. the controls in the MCR. perform active functions.
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2.2.2 12 W Westinghouse oMAACSVev7\it020202.wpf:081298
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l Tier 1 Mitirill l
l PASSIVE CORE COOLING SYSTEM ." g Revision: 7
- Effective: August 1998 2.2.3 Passive Core Cooling System l
Design Description
! The passive core cooling system (PXS) provides emergency core cooling during design basis events.
l The PXS is as shown in Figure 2.2.3-1 and the component locations of the PXS are as shown in Table 2.2.3-5.
- 1. The functional arrangement of the PXS is as described in de Design Description of this Section 2.2.3.
- 2. a) The components identified in Table 2.2.3-1 as ASME Code Section III are designed and constructed in accordance with ASME Code Section III requirements.
b) The piping identified in Table 2.2.3-2 as ASME Code Section Ill is designed and constructed in accordance with ASME Code Section III requirements.
- 3. a) Pressure boundary welds in components identified in Table 2.2.3-1 as ASME Code Section III meet ASME Code Section III requirements.
b) Pressure boundary welds in piping identified in Table 2.2.3 2 as ASME Code Section Ill meet ASME Code Section III requirements.
I 4. a) The components identified in Table 2.2.3-1 as ASME Code Section III retain their pressure boundary integrity at their design pressure. !
b) The piping identified in Table 2.2.3-2 as ASME Code Section III retains its pressure boundary integrity at its design pressure.
- 5. a) The seismic Category I equipment identified in Table 2.2.3-1 can withstand seismic design basis loads without loss of safety function.
b) Each of the lines identified in Table 2.2.3-2 for which functional capability is required is designed to withstand combined normal and seismic design basis loads without a loss of its functional capability.
- 6. Each of the as-built lines identified in Table 2.2.3-2 as designed for leak before break (LBB) meets the LBB criteria, or an evaluation is performed of the protection from the dynamic effects of a rupture of the line.
- 7. a) The Class IE equipment identified in Table 2.2.3-1 as being qualified for a harsh environment can withstand the environmental conditions that would exist before, during, f g and following a design basis accident without loss of rafety function for the time required j Q to perform the safety function.
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2.2.3-1 )
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Tier 1 Mit;ri:1 PASSIVE CORE COOLING SYSTEM f ' Nj Revision: 6 t Effective: July 1998 l-b) The Class IE components identined in Table 2.2.3-1 are powered from their respective Class IE division.
c) Separation is provided between PXS Class IE divisions, and between Class IE divisions and non-Class IE cable.
- 8. The PXS provides the following safety-related functions:
a) The PXS provides containment isolation of the PXS lines penetrating the containment.
b) The PRHR HX provides core decay heat removal during design basis events.
c) The CMTs, accumulators, in-containment refueling water storage tank (IRWST) and containment recirculation provide reactor coolant system (RCS) makeup, boration, and safety injection during design basis events.
d) The PXS provides pH adjustment of water flooding the containment following design basis accidents.
- 9. The PXS has the following features:
a) The PXS provides a function to cool the outside of the reactor vessel during a severe accident.
b) The accumulator discharge check valves (PXS-PL-V028A/B and V029A/B) are of a different check valve type than the CMT discharge check valves (PXS-PL-V016A/B and V017A/B).
I c) The equipment listed in Table 2.2.3-6 has sufficient thermal lag to withstand the effects of I identified hydrogen burns associated with severe accidents.
- 10. Safety-related displays of the parameters identiDed in Table 2.2.3-1 can be retrieved in the main control room (MCR).
I1. a) Controls exist in the MCR to cause the remotely operated valves identified in Table 2.2.3-1 to perform their active function (s).
b) The valves identified in Table 2.2.3-1 as having protection and safety monitoring system (PMS) control perform their active function after receiving a signal from the PMS.
c) The valves identified in Table 2.2.3-1 as having diverse actuation system (DAS) control perform their active function after receiving a signal from the DAS.
- 12. a) The motor-operated and check valves identified in Table 2.2.3-1 perform an active safety-related function to change position as indicated in the table.
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Titr 1 Mit:rirl l
PASSIVE CORE COOLING SYSTEM 3
,q Revision: 6 i
! i
) Effective: July 1998 l
Table 2.2.3-4 I:2spections, Tests, Analyses, and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The functional arrangement of Inspection of the as-built system The as-built PXS conforms with the PXS is as described in the will be performed. the functional arrangement as Design Description of this described in the Design Section 2.2.3- Description of this Section 2.2.3.
2.a) The components identified Inspection will be conducted of The ASME Code Section III in Table 2.2.31 as ASME Code the as-built components as design reports exist for the as-built Section III are designed and documented in the ASME design components identified in constructed in accordance with reports. Table 2.2.3-1 as ASME Code ASME Code Section III Section III.
requirements.
2.b) The piping identified in Inspection will be conducted of The ASME Code Section 111 Table 2.2.3-2 as ASME Code the as-built piping as design reports exist for the as-built Section III is designed and documented in the ASME design piping identified in Table 2.2.3-2 constructed in accordance with reports. as ASME Code Section Ill.
ASME Code Section III requirements.
3.a) Pressure boundary welds in Inspection of the as-built A report exists and concludes that
{~T v/ components identified in pressure boundary welds will be the ASME Code Section III Table 2.2.3-1 as ASME Code performed in accordance with requirements are met for non-Section III rneet ASME Code the ASME Code Section 111. destructive examination of pressure Section III requirements. boundary welds.
3.b) Pressure boundary welds in Inspection of the as-built A report exists and concludes that piping identified in Table 2.2.3 2 pressure boundary welds will be the ASME Code Section III as ASME Code Section III meet performed in accordance with requirements are met for non-ASME Code Section Ill the ASME Code Section III. destructive examination of pressure requirements. boundary welds.
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2.2.3-15 W Westinghouse o:\lT AACS\rev6\it020203.wpf:1 b-071798 u___ . _ _ _ _ _
Tier 1 M3erial PASSIVE CORE COOLING SYSTEM ;.
I Flevision: 7 Effective: August 1998 _
Table 2.2.3-4 (cont.)
Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment Inspections, Tests, Analyses Acceptance Criteria l I
4.a) The components identified A hydrostatic test will be A report exists and concludes that I in Table 2.2.3-1 as ASME Code performed on the components the results of the hydrostatic test of Section III retain their pressure required by the ASME Code the components identified in boundary integrity at their design Section III to be hydrostatically Table 2.2.3-1 as ASME Code pressure. tested. Section III conform with the requirements of the ASME Code Section III.
4.b) 'nie piping identified in A hydrostatic test will be A report exists and concludes that i Table 2.2.3-2 as ASME Code performed on the piping required the results of the hydrostatic test of Section III retains its pressure by the ASME Code Section III the piping identified in boundary integrity at its design to be hydrostatically tested. Table 2.2.3-2 as ASME Code pressure. Section III conform with the requirements of the ASME Code Section III.
5.a) The seismic Category I i) Inspection will be performed i) The seismic Category I equipment identified in to verify that the seismic equipment identified in Table 2.2.3-1 can withstand Category I equipment and valves Table 2.2.31 is located on the seismic design basis loads identified in Table 2.2.3-1 are Nuclear Island.
without loss of safety function. located on the Nuclear Island.
ii) Type tests, analyses, or a ii) A report exists and concludes combination of type tests and that the seismic Category I analyses of seismic Category I equipment can withstand seismic equipment will be performed. design basis dynamic loads without loss of safety function.
iii) Inspection will be performed iii) A report exists and concludes for the existence of a report that the as-installed equipment verifying that the as-installed including anchorage is seismically equipment including anchorage bounded by the tested or analyzed is seismically bounded by the conditions.
tested or analyzed conditions.
5.b) Each of the lines identified Inspection will be performed A report cxists and concludes that in Table 2.2.3-2 for which verifying that the as-built piping each of the as-built lines identified functional capability is required is meets the requirements for in Table 2.2.3-2 for which designed to withstand combined functional capability, functional capability is required normal and seismic design basis meets the requirements for loads without a loss of its functional capability.
functional capability.
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Tier 1 M;tirill MAIN CONTROL ROOM EMERGENCY HABITABILITY SYSTEM Revision: 6 ji j O
Effective: July 1998 Table 2.2.5 6 (cont.)
Component Name Tag Number Component Location Emergency Air Storage Tank 29 VES-MT-29 Auxiliary Building Emergency Air Storage Tank 30 VES-MT-30 Auxiliary Building Emergency Air Storage Tank 31 VES-MT-31 Auxiliary Building Emergency Air Storage Tank 32 VES-MT-32 Auxiliary Building l
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Tier 1 M1 rial MAIN CONTROL ROOM EMERGENCY HABITABILITY SYSTEM .
- 1 Revision: 7 E!!
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Figure 2.2.5-1 Main Control Room Emergency IIabitability System O
2.2.5-16
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F-Tier 1 M;t:ri:1 i
l CONTAINMENT HYDROGEN CONTROL SYSTEM if" i i G Revision: 7 '
i l V Effective: August 1998 l
2.3.9 Containment Hydrogen Control System The containment hydrogen control system (VLS) limits hydrogen gas concentration in containment during accidents.
He VLS has catalytic hydrogen recombiners (VLS-MY-E01 A, VLS-MY-E01B, VLS-MY-E02 and VLS-MY-E03) that are located inside containment. The VLS has hydrogen igniters located as shown on Table 2.3.9-2.
I 1. The functional arrangement of the VLS is as described in the Design Description of this
! Section 2.3.9.
I 2. He seismic Category I equipment identified in Table 2.3.9-1 can withstand seismic design basis loads without loss of safety function.
I 3. a) The equipment identified in Table 2.3.9-1 as being qualified for a harsh environment can withstand the environmental conditions that would exist before, during, and following a design I L
basis accident without loss of safety function for the time required to perform the safety function.
1 b) The Class lE components identified in Table 2.3.9-1 are powered from their respective Class lE division.
c) Separation is provided between VLS Class IE divisions, and between Class lE divisions and 1 j
non-Class IE cable.
I 4. The components identified in Table 2.3.9-2 are powered from their respective non-Class IE power group.
l 5. The VLS provides the following safety-related functions:
a) The VLS provides hydrogen monitors for indication of the containment hydrogen concentration.
b) The VLS provides passive autocatalytic recombiner (PAR) devices for control of the containment hydrogen concentration during and following a design basis accident.
I 6. a) The VLS provides hydrogen monitors for indication of the containment hydrogen concentration, b) The VLS provides PAR devices for control of the containment hydrogen concentration during i- and following a design basis accident.
l 7. The VLS provides the non-safety related function to control the containment hydrogen concentration for beyond design basis accidents. I 2.3.9-1 W85tingh0088 o:\lTAACSVev7\it020309.wpt:1b o81298 u____-_____________. . _ _ _ _ _ _ _ _ _ . _ . _ . . _ _ _ _ _ _
Tier 1 Mit:ri;l j!!!' *-
CONTAINMENT HYDROGEN CONTROL SYSTEM * =
Revision: 7 Effective: August 1998 _
l 8. Safety-related displays identified in Table 2.3.9-1 can be retrieved in the MCR.
I 9. a) Controls exist in the MCR to cause the components identified in Table 2.3.9-2 to perform the listed function.
b) The components identified in Table 2.3.9-2 perform the listed function after receiving a manual signal from the diverse actuation system (DAS).
Inspections, Tests, Analyses, and Acceptance Criteria Table 2.3.9-2 specifies the inspections, tests, analyses, and associated acceptance criteria for the VLS.
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Tier 1 M:t:ri;l PROTECTION AND SAFETY MONITORING SYSTEM ilii
- ^s Revision: 6 L h Effective: July 1998 _
Table 2.5.2-2 PMS Automatic Reactor Trips Source Range High Neutron Flux Reactor Trip Intermediate Range High Neutron Flux Reactor Trip Power Range High Neutron Flux (Low Setpoint) Trip Power Range High Neutron Flux (High Setpoint) Trip Power Range High Positive Flux Rate Trip l Reactor Coolant Pump High Bearing Water Temperature Trip Overtemperature Delta-T Trip Overpower Delta-T Trip Pressurizer Low Pressure Trip i Pressurizer High Pressure Trip l
Pressurizer High Water Level Trip I Low Reactor Coolant Flow Trip Low Reactor Coolant Pump Speed Trip Low Steam Generator Water Level Trip l High-2 Steam Generator Water Level Trip Automatic or Manual Safeguards Actuation Trip Automatic or Manual Depressurization System Actuation Trip k(~)T Automatic or Manual Core Makeup Tank (CMT) Injection Trip i
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PROTECTION AND SAFETY MONITORING SYSTEM
- Revision: 7 j Effective: August 1998 _
j Table 2.5.2-3 PMS Automatically Actuated Engineered Safety Features Safeguards Actuation Containment Isolation Automatic Depressurization System (ADS) Actuation Main Feedwater Isolation Reactor Coolant Pump Trip CMT Injection Turbine Trip Steam Line Isolation i Steam Generator Relief Isolation Steam Generator Blowdown Isolation Passive Containment Cooling Actuation Startup Feedwater Isolation Passive Residual Heat Removal (PRHR) Heat Exchanger Alignment Block of Boron Dilution Chemical and Volume Control System (CVS) Makeup Line Isolation Steam Dump Block MCR isolation and Air Supply Initiation Auxiliary Spray and Letdown Purification Line Isolation Containment Air Filtration System Isolation Normal Residual Heat Removal Isolation Spent Fuel Pool Isolation In-Containment Refueling Water Storage Tank (IRWST) Injection IRWST Containment Recirculation CVS Letdown Isolation Pressurizer Heater Block O
2.5.2-6
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Tier 1 Mrcrial 1
l l PROTECTION AND SAFETY MONITORING SYSTEM fj* Mi
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Revision: 7 f" ill l Effective: August 1998 _
Table 2.5.2-4 i PMS Manually Actuated Functions Reactor Trip Safeguards Actuation Containment Isolation l Stages 1,2, and 3 ADS Actuation Stage 4 ADS Actuation hiain Feedwater Isolation CMT Injection Steam Line Isolation Passive Containment Cooling Actuation PRHR Heat Exchanger Alignment IRWST Injection IRWST Containment Recirculation /lRWST Drain to Containment MCR Isolation and Air Supply Initiation l
Steam Generator Relief Isolation k
1 v
i 2.5.2 7 3 W85tiligh00SP o:VTAACSvev7Vt020502.wpf:1t 081398
Tier 1 M:t: rill PROTECTION AND SAFETY MONITORING SYSTEM i}!! !!!
Revision: 6 -
Effective: July 1998 _ .
Table 2.5.2-5 Alinimum Inventory of Displays, Alerts, and Fixed Position Controls in the A1CR Description Control Display Alert * ,
Neutron Rux -
Yes Yes Neutron Flux Doubling -
No Yes Startup Rate -
Yes Yes Reacter Coolant System (RCS) Pressure -
Yes Yes Wide-range Hot Leg Temperature -
Yes No Wide-range Cold Leg Temperature - Yes Yes RCS Cooldown Rate Compared to the Limit Based on -
Yes Yes RCS Pressure Wide-rar.ge Cold Leg Temperature Compared to the -
Yes Yes Limit Based on RCS Pressure Change of RCS Temperature by more than 5"F in the -
No Yes last 10 minutes Containment Water Level - Yes Yes Containment Pressure -
Yes Yes Pressurizer Water Level - Yes Yes Pressurizer Water Level Trend - Yes No Pressurizer Reference Leg Temperature -
Yes No Reactor Vessel-Hot Leg Water Level -
Yes Yes Pressurizer Pressure - Yes No Core Exit Temperature -
Yes Yes LCS Subcooling -
Yes Yes RCS Cold Overpressure Limit - Yes Yes IRWST Water Level -
Yes Yes PRHR Row -
Yes Yes PRIIR Outlet Tem >erature -
Yes Yes l
l Note: Dash (-) indicates not applicable.
l l 1. These pararneters are used to generate visual alerts that identify challenges to the entical safety functions. For the main control room. the visual alerts are embedded in the safety-related displays as visual signals.
9 2.5.2-8
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Tier 1 Matrial GROUNDING AND UGHTNING PROTECTION SYSTEM ?
Revision: 7
(]
U Effective: August 1998 2.6.6 Grounding and Lightning Protection System l Design Description i
I The grounding and lightning protection system (EGS) provides electrical grounding for ;
1 instrumentation grounding, equipment grounding, and lightning protection during normal and I off-normal conditions.
I i 1. The EGS provides an electrical grounding system for: (1) instrument / computer grounding; I (2) electrical system grounding of the neutral points of the main generator, main step-up I transformers, auxiliary transformers, load center transformers, and onsite standby diesel I generators; and (3) equipment grounding of equipment enclosures, metal stnictures, I metallic tanks, ground bus of switchgear assemblies, load centers, motor control centers, I and control cabinets. Lightning protection is provided for exposed structures and buildings I housing safety-related and fire protection equipment. Each grounding system and lightning )
I protection system is grounded to the station grounding grid. I I
I Inspections, Tests, Analyses, and Acceptance Criteria I
l Table 2.6.6-1 specifies the inspections, tests, analyses, and associated acceptance criteria for
, I the EGS.
U 1 l
I 1
l O
I 2.6.6-1
[ W85tlngh0tlS8 o:vTAACSVev7Voentry.wpf:1two81398 l
Tier 1 Mrrial l GROUNDING AND LIGHTNING PROTECTION SYSTEM ?
i Revislori: 7 Effective: August 1998 l Table 2.6.6-1 1 Inspections, Tests, Analyses, and Acceptance Criteria 1 Design Commitment Inspections, Tc 4.s, Analyses Acceptance Criteria 1 1. The EGS provides an elec.rical i) An inspection for the i) A connection exists between I grounding system for: instrument / computer grounding the instrument / computer I (1) instrument / computer system connection to the station grounding system and the station I grounding; (2) electrical system grounding grid will be performed. grounding grid.
I grounding of the neutral points of I the main generator, main step-up ii) An inspection for the electrical li) A connection exists between I transformers, auxiliary system grounding connection to the electrical system grounding I transformers, load center the station grounding grid will be and the station grounding grid.
I transformers, auxiliary and onsite performed.
I standby diesel generators; and I (3) equipment grounding of iii) An inspection for the iii) A connection exists between I equipment enclosures, metal equipment grounding system the equipment grounding system I structures, metallic tanks, ground connection to the station and the station grounding grid.
I bus of switchgear assemblies, grounding grid will be performed.
I load centers, motor control I centers, and control cabinets. iv) An inspection for the lightning iv) A connection exists between l Lightning protection is provided protection system connection to the lighting protection system and I for exposed structures and the station grounding grid will be the station grounding grid.
I buildings housing safety-related performed.
I and fire protection equipment.
l Each grounding system and I lighting protection system is I grounded to the station grounding i grid.
O E Westirighouse osTAAcsvevnnoentry.wpf:1b- 1 98
Tier 1 Materi:1 NUCLEAR ISLAND NONRADIOACTIVE VENTILATION SYSTEM * !h O
V Revision: 7 d Effective: August 1998 _
2.7.1 Nuclear Island Nonradioactive Ventilation System Design Description The nuclear island nonradioactive ventilation system (VBS) serves the main control room (MCR),
technical support center (TSC), Class lE de equipment rooms, Class lE instrumentation and control I (l&C) rooms, Class IE electrical penetration rooms, Class IE battery rooms, remote shutdown room, reactor coolant pump trip switchgear rooms, adjacent corridors, and the passive containment cooling system (PCS) valve room during normal plant operation. The VBS consists of the following independent subsystems: the main control room / technical support center HVAC subsystem, the class IE electrical room HVAC subsystem and the passive containment cooling system valve room heating and ventilation subsystem. The VBS provides heating, ventilation, and cooling to the areas served when ac power is available. The system pmvides breathable air to the control room and maintains the main control room and technical suppott center areas at a slightly positive pressure with respect to the adjacent rooms and outside environment during normal operations. The VBS monitors the main control room supply air for radioactive particulate and iodine concentrations and provides filtration of !
main control room / technical support center air during conditions of abnormal (high) airbome {
radioactivity. In addition, the VBS isolates the HVAC penetrations in the main control room boundary i on high-high particulate or iodine concentrations in the main control room supply air or on extended loss of ac power to support operation of the main control room emergency habitability system (VES). i
)
(q) The VBS is as shown in Figure 2.7.1-1 and 'he component locations of the VBS are as shown in Table 2.7.1-5.
]
- 1. The functional arrangement of the VBS is as described in the Design Description of this Section 2.7.1.
- 2. a) The components identified in Table 2.7.1-1 as ASME Code Section III are designed and constructed in accordance with ASME Code Section III requirements.
I b) The piping identified in Table 2.7.1-2 as ASME Code Section III is designed and constructed l in accordance with ASME Code Section 111 requirements.
- 3. a) Pressure boundary welds in components identified in Table 2.7.1-1 as ASME Code Section III meet ASME Code Section III requirements.
b) Pressure boundary welds in piping identified in Table 2.7.1-2 as ASME Code Section III meet ASME Code Section III requirements.
l 4. a) The components identified in Table 2.7.1-1 as ASME Code Section III retain their pressure l boundary integrity at their design pressure, i
b) The piping identified in Table 2.7.1-2 as ASME Code Section III retains its pressure boundary integrity at its design pressure.
O r
2.7.1 -1 l T Westinghouse o W AACSVev7\it020701.wpf;1 b-081298
Ti;r 1 M;t:ri;l NUCLEAR ISLAND NONRADIOACTIVE VENTILATION SYSTEM ip Revision: 6 ff Effective: July 1998
- 5. The seismic Category I equipment identified in Table 2.7.1-1 can withstand seismic design basis loads without loss of safety function.
- 6. a) The Class IE components identified in Table 2.7.1-1 are powered from their respective Class lE division.
b) Separation is provided between VBS Class IE divisions, and between Class IE divisions and non-Class 1E cable.
- 7. The VBS provides the safety-related function to isolate the pipes that penetrate the MCR pressure boundary.
- 8. The VBS provides the following nonsafety-related functions:
a) The VBS provides cooling to the MCR, TSC, and Class lE electrical rooms.
b) The VBS provides ventilation cooling to the Class lE battery rooms.
c) The VBS maintains MCR habitability when radioactivity is detected.
d) The VBS provides ventilation cooling via the ancillary equipment in Table 2.7.1-3 to the MCR and the division B&C Class IE I&C rooms.
- 9. Safety-related displays identified in Table 2.7.1-1 can be retrieved in the MCR.
- 10. a) Controls exist in the MCR to cause the remotely operated valves identified in Table 2.7.1-1 to perform their active funct;ons.
b) The valves identified in Table 2.7.1-1 as having protection and safety monitoring system (PMS) control perform their active safety function after receiving a signal from the PMS.
- 11. After loss of motive power, the valves identified in Table 2.7.1-1 assume the indicated loss of motive power position.
- 12. Controls exist in the MCR to cause the components identified in Table 2.7.1-3 to perform the listed function.
- 13. Displays of the parameters identified in Table 2.7.1-3 can be retrieved in the MCR.
Inspections, Tests, Analyses, and Acceptance Criteria Table 2.7.1-4 specifies the inspections, tests, analyses, and associated acceptance criteria for the VBS.
O
[ Westilighouse omTAAcssrevesit020701.wpf;1b- 79
Tier 1 M:t: rial CENTRAL CHILLED WATER SYSTEM f
(
\
Revision: 7 Effective: August 1998 _
2.7.2 Central Chilled Water System Design Description The plant heating, ventilation, and air conditioning (HVAC) systems require chilled water as a cooling medium to satisfy the ambient air temperature requirements for the plant. The central chilled water system (VWS) supplies chilled water to the HVAC systems and is functional during reactor full-power and shutdown operation. The VWS also provides chilled water to selected process systems.
The VWS is as shown in Figure 2.7.2-1 and the component locations of the VWS are as shown Table 2.7.2-3.
I 1. The functional arrangement of the VWS is as described in the Design Description of this l Section 2.7.2.
- 2. The VWS provides the safety-related function of preserving containment integrity by isolation of the VWS lines penetrating the containment.
- 3. De VWS provides the following nonsafety-related functions:
a) De VWS provides chilled water to the supply air handling units serving the MCR, the
)
(j Class lE electrical rooms, and the unit coolers serving the RNS and CVS pump rooms.
b) The VWS air-cooled chillers transfer heat from the VWS to the surrounding atmosphere.
- 4. Controls exist in the MCR to cause the components identified in Table 2.7.2-1 to perform the ;
listed function. !
l I
- 5. Displays of the parameters identified in Table 2.7.2-1 can be retrieved in the MCR.
Inspections, Tests, Analyses, and Acceptance Criteria Table 2.7.2-2 specifies the inspections, tests, analyses, and associated acceptance criteria for the VWS.
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Tirr 1 Mit:ri:1 Iij
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CENTRAL CHILLED WATER SYSTEM Revision: 6 f Effective: July 1998 Table 2.7.21 Control Equipment Name Tag No. Display Function Air-cooled Chiller VWS-MS-02 Yes Start (Run Status)
Air-cooled Chiller VWS-MS-03 Yes Start (Run Status)
Air-cooled Chiller Pump VWS-MP-02 Yes Start (Run Status)
Air-cooled Chiller Pump VWS-MP-03 Yes Start (Run Status)
CVS Pump Room Unit Cooler Fan A VAS-MA-07A Yes Start (Run Status)
CVS Pump Room Unit Cooler Fan B VAS-MA-07B Yes Start (Run Status)
RNS Pump Room Unit Cooler Fan A VAS-MA-08A Yes Start (Run Status)
RNS Pump Room Unit Cooler Fan B VAS-MA-08B Yes Start (Run Status)
Air-cooled Chiller Water Valve VAS-PL-V210 Yes Open (Position Status)
Air-cooled Chiller Water Valve VAS-PL-V253 Yes Open (Position Status)
- gm O1 l 2.7.2-2 l (V Westinghouse o NT AACS\rev6\it020702.wpf:1 b-071498 l
l I
Tier 1 MItirill tlUILDINGS I y i
a Revision: 6 1 Effective: July 1998 I l 3.3 Buildings Design Description i j
i i The nuclear island structures include the containment (the steel containment vessel and the I containment internal structure) and the shield and auxiliary buildings. The containment, shield and auxiliary buildings are structurally integrated on a common basemat which is embedded below the finished plant grade level. The containment vessel is a cylindrical welded steel vessel with elliptical i upper and lower heads, supported by embedding a lower segment between the containment internal structures concrete and the basemat concrete. The containment intemal structure is reinforced concrete j with structural modules used for some walls and floors. The shield building is reinforced concrete i and, in conjunction with the internal structures of the containment building, provides shielding for the reactor coolant system and the other radioactive systems and components housed in the containment.
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l . The shield building roof is a reinforced concrete structure containing an integral, steel lined passive i containment cooling water storage tank. The auxiliary building is reinforced concrete and houses the safety-related mechanical and electrical equipment located outside the containment and shield buildings.
The portion of the annex building adjacent to the nuclear island is a structural steel and reinforced concrete seismic Category 11 structure and houses the technical support center, non-1E electrical equipment, and hot machine shop.
The radwaste building is a steel framed structure and houses the low level waste processing and storage.
" The turbine building is a non-safety related structure that houses the main turbine generator and the power conversion cycle equipment and auxiliaries. There is no safety-related equipment in the turbine !
building. The turbine building is located on a separate foundation. The turbine building structure is l I adjacent to the nuclear island structures.
The diesel generator building is a non-safety related structure that houses the two standby diesel engine powered generators and the power conversion cycle equipment and auxiliaries. There is no safety-related equipment in the diesel generator building. The diesel generator building is located on a ,
l l separate foundation at a distance from the nuclear island structures. !
l The plant gas system (PGS) provides hydrogen, carbon dioxide, and nitrogen gases to the plant systems as required. The component locations of the PGS are located either in the turbine building or the yard areas.
i 1. The physical arrangement of the nuclear island structures and the annex building is as described in
-I the Design Description of this Section 3.3, and as shown on Figures 3.31 through 3.3-14. The I
physical arrangement of the radwaste building, the turbine building, and the diesel generator I building is as described in the Design Description of this Section 3.3.
3.3 1 W85tlligf10ilS8 oNTAACSVev6Nt0303.wpf:073098
Tier 1 M;t:rlil BUILDINGS & .W Revision: 7 =
Effective: August 1998 _
- 2. a) The nuclear island structures, including the critical sections listed in Table 3.3-7, are seismic I Categorf I and are designed and constmeted to withstand design basis loads, as specified in I the Design Description, without loss of structural integrity and the safety-related functions.
The design bases loads are those loads associated with:
. Normal plant operation (including dead loads, live loads, lateral earth pressure loads, and equipment loads, including hydrodynamic loads, temperature and equipment vibration);
. External events (including rain, snow, flood, tornado, tornado generated missiles and earthquake); and
. Internal events (including Good, pipe rupture, equipment failure, and equipment failure generated missiles).
b) Site grade level is located relative to floor elevation 100'-0" per Table 3.3-5. Floor elevation i 100'-0" is defined as the elevation of the floor at design plant grade, c) The containment and its penetrations are designed and constructed to ASME Code Section Ill, Class MC.W d) The containment and its penetrations retain their pressure boundary integrity associated with the design pressure. -
e) The containment and its penetrations maintain the containment leakage rate less than the maximum allowable leakage rate associated with the peak containment pressure for the design basis accident.
f) The key dimensions of the nuclers island stnictures are as defined on Table 3.3-5.
I g) The containment vessel greater than 7 feet above the operating deck provides a heat transfer surface. A free volume exists inside the containment shell above the operating deck.
I 3. Walls and floors of the nuclear island structures as defined on Table 3.3-1, except for designed openings and penetrations, provide shielding during normal operations.
- 4. Walls and floors of the annex building as defined on Table 3.3-1, except for designed openings and penetrations, provide shielding during normal operations.
- 5. a) Exterior walls and the basemat of the nuclear island have a water barrier up to site grade.
b) The boundaries between mechanical equipment rooms and the electrical and instrumentation and control (I&C) equipment rooms of the auxiliary building as identified in Table 3.3-2 are I Containtnent isolation devices are addressed in subsection 2.2.l. Containtnent Systern.
3.3-2 W Westinghouse oMAACSVevM0303.wpf:081298
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Tier 1 M;Orlal l
i BUILDINGS E
.n Revision: 7 "I F l
Effective: August 1998 Table 3.3-5 Key Dimensions of Nuclear Island Building Features Reference Dimension Nominal Key Dimension (Figure 3.314) Dimension Tolerance Distance between Outside Surface of XI 91 ft-0 in +3 ft walls at Column Line I & N when -1 ft Measured at Column Line i Distance from Outside Surface of wall X2 138 ft-0 in +3 ft at Column Line I to Column Line 7 -1 ft when Measured at Column Line I Distance from Outside Surface of wall X3 118 ft-0 in +3 ft at Column Line 11 to Column Line 7 - 1 ft when Measured at Column Line I Distance between Outside Surface of X4 117 ft-6 in +3 ft walls at Column Line I & Q when - 1 ft Measured at Column Line 11 Distance from Outside Surface of wall X5 29 ft-0 in +3 ft g at Column Line Q to Column Line N -1 ft y when Measured at Column Line 11 Distance between Outside Surface of X6 72 ft-6 in +3 ft shield building wall to shield building -1 ft centerline when Measured on West Edge of Shield Building Distance between shield building X7 7 ft-6 in 3 in centerline to Reactor Vessel centerline when Measured along Column Line N in North-South Direction Distance from Bottom of Containment -
2 ft-8 in 2 3 in Sump to Top Surface of Embedded Containment Shell Distance from top of Basemat to Design -
33 ft-6 in I ft Plant Grade l Distance of Design Plant Grade (Floor -
0 ft 3 ft-6 in i elevation 100'-0") relative to Site Grade Distance from Design Plant Grade to -
208 ft-6 in I ft Top Surface of Shield Building Roof 9
3.3-16 W Westinghouse o:MTAACSvev7Vt0303.wpf:081398
l Ti;r 1 MitIrir.1 BUILDINGS q l I "
i Revision: 7 Effective: August 1998 Table 3.3-6 l
Inspections Tests, Analyses, and Acceptance Criteria
! Design Commitment inspections, Tests, Analyses Acceptance Criteria 1.
- l. The physical arrangement of An inspection of the nuclear island The as-built nuclear island l the nuclear island structures and structures, the annex building, the structures, the annex building, the l the annex building is as described radwaste building, the turbine radwaste building, the turbine in the Design Description of this building, and the diesel generator building, and the diesel generator .
I Section 3.3 and Figures 3.3-1 building will be performed. building conform with the through 3.3-14. The physical physical arrangement as described l arrangement of the radwaste in the Design Description of this I building, the turbine building, and Section 3.3 and Figures 3.3-1
{
the diesel generator building is as through 3.3-14.
described in the Design Description of this Section 3.3.
2.a) The nuclear island structures, i) An inspection of the nuclear i) A report exists which including the critical sections island structures will be performed. reconciles deviations during listed in Table 3.3-7, are seismic Deviations from the design due to construction and concludes that Category I and are designed and as-built conditions will be analyzed the as-built nuclear island constructed to withstand design for the design basis loads. structures, including the critical I basis loads as specified in the sections, conform to the approved i Design Description, without loss design and will withstand the j of structural integrity and the design basis loads specified in the i safety-related functions.
l I Design Description without loss of i structural integrity or the safety- l
- related functions. 1 l
ii) An inspection of the as-built ii) A report exists that concludes concrete thickness will be that the as built concrete performed. thicknesses conform with the l building sections defm' ed on l Table 3.3-1.
2.b) Site grade level is located Inspection of the as-built site grade Site grade is consistent with l relative to floor elevation 100'-0" will be conducted. design plant grade within the j l
per Table 3.3-5. dimension defined on Table 3.3-5. j 2.c) The containment and it. See Tier 1 Material, See Tier i Material, l penetrations are designed and Subsection 2.2.1, Containment Subsection 2.2.1, Containment i constructed to ASME Code System. System. i Section III, Class MC.W l r
L l
l
- 1. Containment isolation devices are addressed in subsection 2.2.1, Containment System.
3.3-17 W85tiflgh00S8 oNTAACS\rev7\it0303.wpf:081298
Ti;r 1 M;t;ri:1 BUILDINGS iin y Revision: 7 9 Effective: August 1998 _
Table 3.3-6 (cont.)
Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment Inspections, Tests, Arralyses Acceptance Criteria 2.d) The c$minment and its See Tier 1 Material, Subsection See Tier 1 Material, Subsection penetrations retain their presc 2.2.1, Containment System. 2.2.1, Containment System.
boundary integrity associated with the design pressure.
2.e) The containment and its See Tier 1 Material, Subsection See Tier 1 Material, Subsection penetrations maintain the 2.2.1, Containment System. 2.2.1, Containment System.
containment leakage rate less than the maximum allowable leakage rate associated with the peak containment pressure for the design basis accident.
2.f) The key dimensions of An inspection will be performed of A report exists and concludes that nuclear island structures are the as-built configuration of the the key dimensions of the as-built defined on Table 3.3-5. nuclear island structures. nuclear island structures are consistent with the dimensions defined on Table 3.3-5.
I 2.g) The containment vessel The maximum containment vessel The containment vessel maximum greater than 7 feet above the inside height from the operating inside height from the operating operating deck provides a heat deck is measured and the inner deck is 121'-1" (with tolerance of transfer surface. A free volume radius below the spring line is + 12", -6"), and the inside diameter exists inside the containment shell measured at two orthogonal radial is 130 feet nominal (with above the operating deck. directions at one elevation. tolerance of + 12", -6").
- 3. Walls and floors of the nuclear Inspection of the as-built nuclear A report exists and concludes that I island structures as defined on island structures wall and floor the shield walls and floors of the i Table 3.31 except for designed thicknesses will be performed. nuclear island structures as openings or penetrations provide defined on Table 3.3-1 except for shielding during normal operations. desiFned openings or penetrations are consistent with the concrete wall thicknesses provided in Table 3.3-1.
O 3.3 18 T Westinghouse o:\lTAACS\rev7\it0303.wpf:081298 1
Tier 1 M tirill ut :
BUILDINGS 9
Revleion: 7 4 Effective: August 1998 _
Table 3.3-6 (cont.)
Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 4. Walls and Doors of the annex Inspection of the as-built annex A report exists and concludes that building as defined on building wall and floor the shield walls and floors of the l Table 3.3-1 except for designed thicknesses will be performed. annex building as defined on openings or penetrations provide Table 3.3-1 except for designed shielding during nonnal openings or penetrations are operations, consistent with the ;ninimum concrete wall thicknesses provided i in Table 3.3-1.
5.a) Exterior walls and the An inspection of the as-built A report exists that confirms that a basemat of the nuclear island exterior walls and the basemat of water barrier exists on the nuclear have a water banier up to site the nuclear island up to floor island exterior walls up to site grade, elevation 100'-0", for application grade, of water barrier will be performed j during construction before the walls are poured.
l l l 5.b) The boundaries between An inspection of the auxiliary A report exists that confirms Doors 1 rooms identified in Table 3.3-2 of building rooms will be and walls as identified on l
y the auxiliary building are performed. Table 3.3-2 have provisions to l
j designed to prevent flooding of prevent Gooding between rooms up
- rooms that contain safety-related ' to the maximum Good levels for
- equipment. each room defined in Table 3.3-2.
5.c) The boundaries between the An inspection of the boundaries A report exists that confirms that
- following rooms, which contain between the following rooms provisions to prevent flooding of safety-related equipment - PXS which contain safety-related other rooms to a maximum Coor l
l valve / accumulator room A equipment - PXS Valve / level of 108 feet are provided.
l (11205), PXS valve / accumulator Accumulator Room A (11205),
l room B (11207), and CVS room PXS Valve / Accumulator Room B (11209)- are designed to prevent (11207), and CVS Room !
flooding between these rooms. (11209)- will be performed. !
l 6.a) The available room volumes An inspection will be performed A report exists and concludes that of the radiologically controlled of the as-built radiologically the as-built available room volumes area of the auxiliary building controlled area of the auxiliary of the radiologically controlled area between Door elevations 66'-6" building between Door elevations of the auxiliary building between and 82'-6" exceed the volume of 66'-6" and 82'-6" to define Door elevations 66'-6" and 82'-6" the liquid radwaste storage tanks volume, exceed the volume of the liquid (WLS-MT-OSA, MT-05B, radwaste storage tanks MT-06A, MT-%B, MT-07A, (WLS-MT-05A, MT-05B, MT-06A, MT-07B, MT-07C, MT-11). MT-06B, MT-07A. MT-07B, MT-07C, MT il).
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3.3-19 MSHngh0LIS8 oAITAACS\rev7Vt0303.wpf:081298
l Tl;r 1 Mit: rill 1
BUILDINGS P*
Revision: 6 Effective: July 1998 _
Table 3.3-6 (cont.)
Inspections, Tests, Analyses, and Acceptance Criteria Design Committnent Inspections, Tests, Analyses Acceptance Criteria 6.b) The radwaste buildmg An inspection of the radwaste The volume of the radwaste package waste storage room has a building packaged waste storage building packaged waste storage volume greater than or equal to room (50352) is performed. room (50352) is greater than or 1293 cubic feet. equal to 1293 cubic feet.
l 7.a) Class lE electrical cables, Inspections of the as-built Class lE electrical cables, fiber I fiber optic cables associated with Class lE cables and raceways optic cables associated with only I only one division, and raceways will be conducted. one division, and raceways are are identified according to identified by the appropriate color applicable color-coded Class IE code.
divisions.
I 7.b) Class IE divisional electrical Inspections of the as-built Class lE electrical cables and fiber I cables and fiber optic cables Class lE divisional cables and optic cables associated with only I associated with only one division raceways will be conducted. one division are routed in raceways are routed in their respective assigned to the same division.
divisional raceways. There are no other safety division i electrical cables in a raceway assigned to a different division.
7.c) Separation is mairtained i) Inspections of the as-built i) Results of the inspection will I between Class IE divisions in Class lE division electrical confirm that the separation between I accordance with the fire areas as cables, fiber optic cables Class IE divisions is consistent with I identified in Table 3.3-3. associated with only one division, Table 3.3-3.
and raceways located in the fire areas identified in Table 3.3-3 will be conducted.
ii) Inspections of the as-built fire ii) Results of the inspection will barriers between the fire areas confirm that fire barriers exist identified in Table 3.3-3 will be between Class IE divisions conducted. consistent with the fire areas identified in Table 3.3-3.
O l 3.3-20 T Westinghouse a:uTAACS\rev6Vt0303.wpf:073198
. 1 Tiir 1 M tirill BUILDINGS C
- Revision: 7
- 4 L Effective: August 1998 Table 3.3-6 (cont.)
Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria 7.d) Physical separation is Inspections of the as-built Results of the inspection will maintained between Class IE Class IE raceways will be confirm that the separation between
{
divisions and between Class IE performed to confirm that the Class lE raceways of different '
divisions and non-Class IE separation between Class IE divisions and between Class lE l cables. raceways of different divisions raceways and non-Class lE and between Class 1E raceways raceways is consistent with the l and non-Class IE raceways is followings:
l consistent with the following:
l
- Within the main control room - Within the main control room i and remote shutdown area, and remote shutdown area, the the minimum vertical vertical separation is 3 inches or separation is 3 inches and the more and the horizontal minimum horizontal separation is 1 inch or more.
separation is 1 inch.
- Within other plant areas - Within other plant areas (limited (limited hazard areas), the hazard areas), the separation minimum separation is meets one of the following:
i defined by one of the
\ following:
- 1) The minimum vertical 1) The vertical separation is !
I separation is 5 feet and 5 feet or more and the the minimum horizontal horizontal separation is separation is 3 feet. 3 feet or more except.
- 2) The minimum vertical 2) The minimum vertical l l separation is 12 inches separation is 12 inches and
- and the minimum the minimum horizontal l horizontal separation is separation is 6 inches for 6 inches for raceways raceweys containing only containing only instrumentation and control instrumentation and and low-voltage power ;
control and low-voltage cables <2/0 AWG. l l power cables <2/0 AWG.
I 7.e) Class IE fiber optic cables Inspections of the as-built Class IE fiber optic cables which I which interconnect two divisions Class lE fiber optic cables will interconnect two divisions are i are routed and separated such that be conducted, routed and separated such that the I the Protection and Safety Protection and Safety Monitoring l Monitoring System voting logic is System voting logic is not defeated I not defeated by the loss of any by the loss of any single raceway or I single raceway or fire area. fire area.
O 3.3-21 W85tifgh0LIS8 o:MTAACSVev7Vt0303.wpf:081298
Ti:r 1 M;t:ri:1 BUILDINGS :"' f6
" J Revision: 6 Effective: July 1998 _
Table 3.3-6 (cont.)
Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 3) For configurations that 3) For configurations that involve exclusively involve exclusively limited limited energy content energy content cables cables (instrumentation (instrumentation and and control), the control), the minimum minimum vertical vertical separation is separation is 3 inches 3 inches and the minimum and the minimum horizontal separation is horizontal separation is 1 inch.
1 inch.
- 4) For configurations 4) For configurations that insolving an enclosed involve an enclosed raceway and an open raceway and an open raceway, the minimum raceway, the minimum vertical separation is vertical separation is 1 inch I inch if the enclosed if the enclosed raceway is raceway is below the below the raceway.
open raceway.
- 5) For configuration 5) For configurations that involving enclosed involve enclosed raceways, raceways, the minimum the minimum vertical and separation is 1 inch in horizontal separation is both horizontal and 1 inch.
vertical directions.
- Where minimum separation - Where minimum separation distances are not maintained, distances are not met, the the circuits are run in circuits are run in enclosed enclosed raceways or barriers raceways or barriers are are provided. provided.
- Separation distances less than - A report exists and concludes those specified above and not that separation distances less run in enclosed raceways or than those specified above and provided with barriers are not provided with enclosed based on analysis, raceways or barriers have been analyzed.
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TIIr 1 M trri;l BUILDINGS . r 1 Revision: 6 i si
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Table 3.3-6 (cont.)
Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 12. The extended turbine An inspection of the as-built The extended axis of the turbine generator axis intersects the shield turbine generator will be generator intersects the shield building. performed. building.
I 13. Separation is provided An inspection of the separation of The minimum horizontal clearance i between the structural elements of the nuclear island from the annex, above Door elevation 100'-0*
I the turbine, annex and radwaste radwaste and turbine building between the structural elements of I l buildings and the nuclear island structures will be performed. The the annex and radwaste buildings I structure. This separation permits inspection will verify the and the nuclear island is 4 inches.
I horizontal motion of the buildings specined horizontal clearance The minimum horizontal clearance i l in the safe shutdown earthquake between structural elements of the above floor elevation 100'-0" I without impact between structural adjacent buildings, consisting of between the structural elements of I elements of the buildings, the reinforced concrete walls and the turbine building and the nuclear i slabs, structural steel columns and island is 12 inches.
I floor beams, l 14. Protected Area / Vital Area An inspection of the as-built The as-built inspection report exists l l walls that are accessible and Protected Area / Vital Area walls and concludes that the Protected l l unmonitored are security that are accessible and Area / Vital Area walls that are
! I hardened. unmonitored will be performed. accessible and unmonitored meet the OI requirements of being security V l hardened,
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Ti;r 1 M;t; rill BUILDINGS Revision: 7 Effective: August 1998 Table 3.3-7 Nuclear Island Critical Structural Sections Containment Internal Structures South west wall of the refueling cavity Sc ith wall of the west steam generator cavity east wall of the in-containment refueling water storage tank In- ;-a -' refueling water storage tank steel wall Colum.. . ., . ig the operating floor Auxiliary and Shield Buildine South wall of auxiliary building (column line 1), elevation 66t6" to elevation 18050" Interior wall of auxiliary building (column line 7.3), elevation 66'-6" to elevation 160'-6" West wall of main control room in auxiliary building (column line L), elevation 117t6" to elevation 153'-0" North wall of MSIV east compartment (column line 11 between lines P and Q), elevation 117*-6" to elevation 153*-0" Shield building cylinder, elevation 160i6" to elevation 200'-0" Roof slab at elevation 180iO" adjacent to shield building cylinder Floor slab on metal decking at elevation 135t3" 2t0" slab in auxiliary building (tagging room ceiling) at elevation 135'-3" Finned floor in the main control room at elevation 135'-3" l Shield building roof, exterior wall of the PCS water storage tank l Shield building roof, tension ring and columns between air inlets, elevation 24110" to elevation 250'-0" Divider wall between the spent fuel pool and the fuel transfer canal Nuclear Island Basemat Below Auxiliary Building Bay between reference column lines 9.1 and i1, and K and L Bay between reference column lines 1 and 2 and K-2 and N O
T Westinghouse o:\lTAACS\rev7\it0303.wpt.08139
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Figure 3.3-14 Nuclear Island Structures Dimensions at Elevation 66'-6" 3.3-53 0:\ap600VTAACSVev7Mt030314.wpf:1b-081398
l Tier 1 M;terial RADIATION MONITORING g.
( Revleion: 7 Effective: August 1998 3.5 Radiation Monitoring Design Description 1 Radiation monitoring is provided for those plant areas where there is a significant potential for
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l airbome contamination, for those process and effluent streams where contamination is possible, and in I accessible areas to provide indication of unusual radiological events as identified in Tables 3.5-1, 3.5-2,3.5-3,3.5-4, and 3.5-5. The radiation monitoring component locations are as shown in Table 3.5-7.
- 1. The seismic Category I equipment identified in Table 3.5-1 can withstand seismic design basis loads without loss of safety function.
- 2. The Class IE equipment identified in Table 3.5-1 as being qualified for a harsh environment can withstand the environmental conditions that would exist before, during, and following a design basis accident without loss of safety function for the time required to perform the safety function.
- 3. Separation is provided between system Class IE divisions, and between Class IE divisions and non-Class IE cable.
- 4. Safety-related displays identified in Table 3.5-1 can be retrieved in the main control room (MCR).
- 5. The process radiation monitors listed in Table 3.5-2 are provided.
- 6. The effluent radiation monitors listed in Table 3.5-3 are provided.
- 7. The airbome radiation monitors listed in Table 3.5-4 are provided.
- 8. The area radiation monitors listed in Table 3.5-5 are provided.
Inspections, Tests, Analyses, and Acceptance Criteria Table 3.5-6 specifies the inspections, tests, analyses, and associated acceptance criteria for radiation monitoring.
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O 3.5-1 MM 0:Vp01\newproj2VTAACS\rev7Vt0305,wpf:081398 L
Tier 1 Mit';ri;l FIADIATION MONITORING jF i Revision: 6 Effective: July 1998 _
l Table 3.5-1 Qual.
for Safety-Seismic Ilarsh Related Equipment Name Tag No. Cat. I Class IE Envir. Display Containment High Range PXS-RE160 Yes Yes Yes Yes Monitor {
Containment High Range PXS-RE161 Yes Yes Yes Yes Monitor Containment High Range PXS-RE162 Yes Yes Yes Yes Monitor Containment High Range PXS-RE163 Yes Yes Yes Yes Monitor MCR Radiation Monitoring VBS-RE01 A Yes Yes No Yes Package A")
MCR Radiation Monitoring VBS-RE01B Yes Yes No Yes Package B")
Containment Atmosphere PSS-RE026 Yes No No No Monitor (Gaseous)
Containment Atmosphere PSS-RE027 Yes No No No Monitor (N13)
Notes: (1) Each MCR Radiation Monitoring Package includes particulate, iodine and gaseous radiation monitors.
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SITE PARAMETERS jg
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Table 5.0-1 (cont.)
Site Parameters f Soil Average Allowable Static Greater than or equal to 8,000 lb/ft2 over the footprint of the nuclear ,
Soil Bearing Capacity island at its excavation depth Lateral Variability Soils supporting the nuclear island should not have extreme variations in subgrade stiffness.
Case 1: For a layer with a low strain shear wave velocity greater than or equal to 2500 feet per second, the layer should have approximately uniform thickness, should have a dip not greater than 20 degrees, and should have less than 20 percent variation in the shear wave velocity from the average velocity within any layer.
Case 2: For a layer with a low strain shear wave velocity less than 2500 feet per second, the layer should have approximately uniform thickness, should have a dip not greater than 20 degrees, and should have less than 10 percent variation in the shear wave velocity from the average velocity within any layer, t
l Q,O Shear Wave Velocity Greater than or equal to 1000 ft/sec based on low-strain, best-estimate soil propernes Liquefaction Potential None Seismic Safe Shutdown SSE free field peak ground acceleration of 0.30 g at plant grade with l Earthquake (SSE) modified Regulatory Guide 1.60 response spectra (see Figures 5.0-1 and l 5.0-2) and the response spectra shown in Figures 5.0-3 and 5.0-4 at the l
l foundation level 40 feet below Design Plant Grade.
Fault Displacement None Potential Atmospheric Dispersion Factors (X/Q)
Site Boundary 0- to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time interval si.0 x 10-3 sec/m 3 l
Annual average $2.0 x 10-5 sec/m 3 Low Population Zone 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> $1.35 x 104 sec/m 3 Boundary 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 51.0 x 10d sec/m 3 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 55.4 x 10-5 sec/m 3 96 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 52.2 x 10-5 sec/m 3 O
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4 Ti:r 1 Mit: rill SITE PARAMETERS M *t Revision: 6 -
Effective: July 1998 l
AP600 Hqrizontal Design Response Spectra 2.8 . ,
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Figure 5.01 IIorizontal Design Response Spectra Safe Shutdown Earthquake O
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