NSD-NRC-97-5137, Forwards Formal Transmittal of Correspondence Previously Sent Informally Over Period of 970416-30,including Info Re SSAR Open Items 292,277,300 & 304
| ML20141K118 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 05/20/1997 |
| From: | Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD-NRC-97-5137, NUDOCS 9705280325 | |
| Download: ML20141K118 (127) | |
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4 Westinghouse Energy Systems sa ass
' Electric Corporation Fstisburgh Pennsylvania 15230-0355 NSD-NRC-97-5137 DCP/NRC0876 Docket No.: STN-52-003 May 20,1997 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: T. R. QUAY
SUBJECT:
INFORMAL CORRESPONDENCE
Dear Mr. Quay:
Please find enclosed a formal transmittal of correspondence we have previously sent to you informally.
This informal correspondence was sent over the period April 16,1997 through April 30,1997. provides the index of the attached material as you have requested.
J'/
.d Brian A. McIntyre, Mana s Advanced Plant Safety and Licensing i
jml Attachment Enclosure cc:
N. J. Liparulo, Westinghouse (w/o Attachment, Enclosure)
M. M. Slosson, NRC (w/o Enclosure) l s
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nua 9705280325 970520 PDR ADOCK 05200003 A
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Attachment I to Westinghouse Letter DCP/NRC0876 DATE ADDRESSEE DESCRIPTION 4/16/97 Scaletti Information related to open item 292. Was originally submitted to NRC in our letter of 3/5/97 and fax to Diane Jackson on 4/10/97. Request NRC' review material and provide definitive action or provide direction to change NRC status to
" Action N" or " closed".
4/16/97 Scaletti information related to open item 277. Was originally submitted to NRC in revision 7 of the SSAR (5/6/96). Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or " closed".
4/16/97 Scaletti information related to open item 298. Was originally submitted to NRC in revision 11 of the SSAR (2/28/97 and our letter of 3/5/97. Request NRC review material and provide dermitive action or provide direction to change NRC status to
" Action N" or " closed" 4/16/97 Scaletti information related to open item 300. Was originally submitted to NRC in fax to Diane Jackson on 1/9/97 and SSAR revision 11 (2/11/97. Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or " closed".
4/16/97 Scaletti information related to open item 304. Was originally submitted to NRC in revision 7 of the SSAR (4/30/96) and a fax to Diane Jackson on 1/28/97. Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or " closed" 4/23/97 11uffman Draft responses to NRC letters on ASI and ERG. Request comments.
4/26/97 Bongarra Draft tier 1 material for minimum inventory.
4/25/97 Bongarra Draft tier 1 material for emergency response facilities. Request to comment by 5/6/97.
4/28/97 Jackson Copy of Westinghouse letter of 4/25/97 concerning formal notification of resolution of open items for SSAR section 9.4.
4/29/97 Iluffman Drafl response to Q4 from NRC letter of 10/3/96. Also, sample of mode 4 tech spec.
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i 4/28/97 Scaletti information related to open item 1102. SRP section 9.3.4 does not apply since CVS is not safety related.
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4/28/97 Scaletti Information related to open item 1101. Since CVS is not RTNSS important, this item can be closed. Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or " closed" 4/29/97 Scaletti information related to open item 3088. Information originally supplied in Westinghouse letter of 7/1/96. Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or " closed" 4/29/97 Scaletti information related to open item 1112. Information originally supplied in fax to Jackson of 1/28/97 and SSAR revision 7 (4/30/96). Other related open items are closec.
Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or "clored".
4/28/97 Scaletti Information related to open item 3085. Infonnation -
nally supplied in Westinghouse letter of 5/20/96. Reque review material and provide dermitive action or provid, direction to change NRC status to " Action N" or "close:
Information related to open item 1458. Information original [y 4/28/97 Scaletti supplied in revision 8 of the SSAR (7/96).
Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or " closed".
4/28/97 Scaletti Information related to open item 1176. Information requested contained in section 1.8 of the SSAR. Requt;st NRC review material and provide dermitive action or provide direction to change NRC status to " Action N" or " closed".
4/30/97 Bongarra Draft markup of WCAP 14401 to be consistent with dran ITAAC. Request comments by 5/6/07.
4/17/97 iluffrnan Comments on draft meeting minutes of 3/13/97 NOTRUMP meeting.
4/28/97 Scaletti information related to open items 1888,1975 and 3271.
Information originally supplied in revision 7 of the SSAR (5/6/96). Additional infonnation supplied in our letter of 2/11/97. Request NRC review material and provide definitive action or provide direction to change NRC status to " Action N" or " closed" 32iu
b FAX to DINO SCALETTI April 16,1997 CC:
Sharon or Dino, please make copies for:
Diane Jackson Ted Quay Don Lindgren Don Hutchings Bob Vijuk Brian McIntyre OPEN ITEM #292 (M9.4.8-1)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 45 calendar days away (33 business days). The relevant documentation related to Open item #292 (M9.4.8-1) is in our letter to NRC
' # NSD-NRC-5012 sent to you March 5,1997 and a fax from D. Lindgren to D. Jackson on April 10 which transmitted a draft change page for Revision 12 of SSAR Table 3.2-3 (Sheet 58 of 64).
Pertinent pages of these documents are attached. It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this item.
We recommend " Action N" or " Closed."
Jim Winters 412-374-5290 l
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.' AP688 Open Itene Tracking Systeam Databese: Executiv2 Summary Data: #16/97 Selectese:
btem noj between 292 And 292 Sorted by item #
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lion DSER Section Tule/ Description Resp (w)~
.NRC No
. Branch Questson Type Detad Status -
Engineer Status Status Letter No. I Dane
~292
. NRR/SPLB 9.48 MiGOI 32-3/Lindgren Action N Action W lM9 4 8-I (REDWASTE BlJILDING IIEAC SYSTEM)
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[ Revise Table}2-3 of the SSAR for vacuum relief syseem (VRS) components including unlisted system dampers and high and low efficiency filects
[
, ICE ITEUahdBuildangilVAC System (VRS) system EU have components that are cI5 fed'as AP60U Equipment Cidli,2,7 I
D. Only Class A, B,C, and D eqmpment and valves are included on Table 3.2-3. The VRS is not included in Table 3.2-3 1
.t Action W - this is the same as OITS8 285 for VRS.
Action N - Response presided by NSD-NRC-97-5012 of 3/5/97 and a fan dased 4/I087. jww-l e
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l Westinghouse Energy Systems 80: 355 l
Electric Corporation Pmsburgh Pennsywama 15230-0355 i
NSD-NRC-97-5012 DCP/NRC076 Docket No.: STN-52-003 March 5,1997 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 i
ATTENTION: T.R. QUAY
SUBJECT:
RESPONSE TO RAI 410.295 CLASSIFICATION OF MECHANICAL AND FLUID SYSTEMS
Dear Mr. Quay:
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Attact;ed are responses to NRC RAI 410.29 and a related open item (OITS #289). Also attached is a copy of SSAR subsection 3.2.4 that shows the changes made in Revision 11 that implements the commitment in the response. Table 3.2-2 has been revised in SSAR Revision 11 to include all fluid and mechanical systems, identify the classification for all components, and provide classification of fire dampers. Table 3.2-3 has also been revised in SSAR Revision 11 to include the changes due to other responses and design changes identified to date.
This response will permit Plant Systems Branch to close these items and finalize its input to FSER i
section 3.2.
.i If you have any questions please contact D. A. Lindgren at (412) 374-4856.
f/f Brian A.
cintyre, Manager Advanced Plant Safety and Licensing jml i
Attachment cc:
Diane Jackson, NRC (w/ Attachment) l W. Huffman, NRC (w/ Attachment)
=_ O
- 3. Design of Structures, Componen^a, Equipment and Systems c
Table 3.2-2 SEISMIC CLASSIFICATION OF BUILDING STRUCTURES Structure Category Nuclear Island C-I Basemat Containment Interior Shield Building Auxiliary Building Containment Air Baffle Containment Vessel C-I Plant vent and stair structure C-II
- 1..
Turbine Building NS Annex Building Columns A - F NS
_ Anlldin, Cuiumns F - I
-M A
e Diesel-Generator Building
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Radwaste Building NS NS Circulating Water Pumphouse and Towers NS C-I-Seismic Category I C-II-Seismic Category II NS-Non-r.eismic Note:
1.
Within the broad definition of seismic Category I and II structures, these buildings contain members and structural subsystems the failure of which would not impair the capability for safe shutdown. Examples of such systems would be elevators, stairwells not required for access in the event of a postulated earthquake, and nonstructural partitions in nonsafety-related areas. These substructures are classified as non-seismic.
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9 Ld Revision: 11 February 28,1997 3.2-20 3 Westingh0US8
A Attachment to NSD-NRC-97 5012 RAI# 410.295 (3481)- NRC Letter 8/15/1996 SSAR Table 3.2-3, AP600 Classification of Mechani-cal and Fluid Systems, Components, and Equipment:
Westinghouse needs to revise Table 3.2-3 to provide the classification of the following fluid
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a.
systems and their associated generalized equipment:
1.
Radiologically. Controlled Area Ventilation System (VAS)
- 2. Containment Recirculation Cooling System (VCS)
- 3. Health Physics and Hot Machine Shop HVAC system (VHS)
- 4. Radioactive Waste Building HVAC System (VRS)
- 5. Turbine Building Ventilation System (VTS)
- 6. Annex / Auxiliary Nonradioactive Ventilation System (VXS)
- 7. Liquid Waste Management System
- 8. Gaseous Waste Management System
- 9. Radiation Monitoring System
- 10. Main Steam System
- 11. Condensate Storage System
- 12. Reactor Coolant Pressure Boundary (RPCB) Leakage Detection and Monitoring System b.
In the previous version, there was a " Location" colunm in the table, which is useful to the reviewer. It was removed from the table in Revision 8. Bring the location information back to the table.
Westinohnuse Resnon=> 7 Table 3.2-3 focuses on the classification of safety-related (Class A, B, or C) components and a.
equipment in mechanical and fluid systems. Items that are AP600 equipment Class D or equivalent are identified in a general basis. The table is being revised to include the fluid and mechanical systems in AP600. Items that are not Class A, B, C, or D are not individually identified. Systems that are electrical or instrumentation systems are not included in this table.
This is consistent with the guidance in Regulatory Guide 1.70. The components in the incore instrumentation system that have a pressure boundary function ars isluded inthe tab Responses for the specific systems follows. The text in subsection 3.2.4 referencing the contents of the table has also been revised.
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- 1. The radiologically controlled area ventilation system (VAS) is included in the table. The Class D room coolers and valves that provide a Class D function are identified. The balance of the equipment in the VAS is Class E.
- 2. The equipment in the containment recirculation cooling system (VCS) is Class E or Class L.
- 3. The equipment in the health ohy=i" ~.3 b -Wa shop HVAC system (VHS) is Class E.
- 4. The equipment in the radioactive waste building HVAC System (VRS) is Class E, Class L, Class F, or Class R.
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,e Attachment to NSD-NRC-97-5012 3.2.4 Application of AP600 Safety-Related Equipment and Seismic Classification System The application of the AP600 equipment and seismic classification system to AP600 systems and components is shown in Table 3.2-3. Table 3.2-3 lists safety-related and seismic category I mechanical and fluid system component and associated couipment class and seismic category as well as other related information. The table also provides information on the systems that I
contain Class D components. Additional information on the Class D functions of the various
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- I systems can be found in the description in the SSAR for the systems. Mechanical and fluid I-systems that contain no safety related or Class D systems are included in the table and general I
information provided on the system. Supports for piping'and components have the same j
classification as the component or piping supported. Supports for AP600 equipment Class A, B, and C mechanical components and piping are constructed to ASME Code,Section III, Subsection NF requirements. The principle construction code for supports for nonsafety-related components and piping is the same as that for the supported component or piping.
I Following the name of each system is the building location of the system components. Some of 1
the systems supply all or most of the buildings. This is indicated by identifying the location as l
various. - Where a system includes piping or ducts that only passed through a building without I
including any components that building is generally not included in the list:
g The following list includes the systems in Table 3.2-3. The three letters in the beginning of each.
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line is the acronym for the system. The systems included in Table 3.2 3 are listed alphabetically (
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l by three letter acronym. TM :j:::= ; ':::d !: fr!!r: br: :: :: p::::: != C! rr A, B, C.
I
= D d br: :: !!:: ef ::=p:::: '!=!M S 1: d':. Those systems marked with an I
l asterisk
- are electrical or instrumentation systems and are not included in Table 3.2-3. The I
components in the incore instrumentation system that have a pressure boundary function are
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l included in the table. See Section 3.11 for identification of safety-related electrical and instrumentation equipment.
m NSSS/ Steam Generator Controls Ami Auxiliade BDS
. Steam Generator Blowdown System CNS Containment System CVS Chemical and Volume Control System PCS Passive Containment Cooling System PXS Passive Core Cooling System RCS Reactor Coolant System RNS Normal Residual Heat Removal System RXS -
Reactor System SGS Steam Generator System Nuclear control and Monitodng i
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- DAS Diverse Actuation System I
- IIS Incore Instrumentation System
- OCS Operation and Control Centers
- PMS Protection and Safety Monitoring System PSS Primary Sampling System
- RMS Radiation Monitoring System S
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Attachment to NSD-NRC-97-5012 i
VHS
_ Health-PhyEnd Hot MachiTil!IlioirHVACJ stem l
VLS Containment Hydrogen Control System s
VRS Radwaste Building HVAC System VTS Turbine Building Ventilation System l
VUS Containment Leak Rate Test System QWS Central Chilled Water System VXS'- AnnexlAuxiliary NQaradinactwe Ventih!!c. Spmu 1
VYS Hot Water Heating System' VZS
- Diesel Generator Building Ventilation System d
Turbine-Generator Controls and Auvillary CMS Condenser Air Removal System UCS Generator Hydrogen and CO Systems 2
HSS Hydrogen Seal Oil System LOS Main Turbine and Generator Lube Oil System
- TOS Main Turbine Control and Diagnostics System Material Handling
'FHS Fuel Handling and Refueling System MHS Mechanical Handling System i
Piping Services CAS Compressed and Instrument Air Systems DOS Standby Diesel and Auxiliary Boiler Fuel Oil System FPS Fire Protection System PGS Plant Gas Systems PWS Potable Water System Non-Class 1E Power Systems i
- ECS Main AC Power System i
- EFS
- Communication Systems
- EGS Grounding and Lightning Protection System
- EHS Special Process Heat Tracing System i
- ELS Plant Lighting System
- EQS Cathodic Protection System Non-Nuclear Controls and Monitoring
- DDS Data Display and Processing System j
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- MES Meteorological and Environmental Monitoring System i
- PLS Plant Control System
- SES Plant Security System SSS Secondary Sampling System
'TVS-Closed Circuit TV System 7N mm.
S
IM Westinghouse FAX COVER SHEET D
1 RECIPIENT INFORMATION SENDER INFORMATION DATE:
4 /f o /9 7 NAME:
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LOCATION:
ENERGY CENTER -
. I Ac.a<ow EAST 3 3e, PHONE:
FACSIMILE:
PHONE:
Omce: (9,y 3 >4-4 8d COMPANY:
Facsimile:
win:
284-4887 NA(
outside: (412)374-4887 LOCATION:
g e,cf g, g g Cover + Pages 1+
t The following pages are being sent from the Westinghouse Energy Center, East Tower, l
Monroeville, PA. If any problems occur during this transmission, please call:
WIN: 284-5125 (Janice) or Outside: (412)374-5125.
COMMENTS:
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- 3. Design of Structures, Coluponents, Equipment, and Systems Table 3.2 3 (Sheet 58 of 64)
AP600 CLASSIFICATION OF MECHANICAL AND FLUID SYSTEMS, COMPONENTS, AND EQUIPMENT Tag Number Description AP600 Seismic Principal Con-Comments Class Category struction Code Containment Hydrogen Control System (VLS)
Location: Containment n/a Hydrogen Igniters D
NS Manufacturer Provides Std.
Hydrogen Control Following Severe Accidents VLS-MY-E01 A Catalytic Hydrogen C
1 Manufacturer Recombiner A Std.
VLS MY-E01B Catalytic Hydrogen C
1 Manufacturer Recombiner B Std.
n/a Fire Dampers Note 3 NS UL 555 Balance of system components are Class E or Class L l
Radwaste Building Ventilation System (VRS)
Location: Radwaste Building -
l n/a Shutoff, Isolation, and L
NS ANSI /AMCA-l Balancing Dampers 500 l
n/a Fire Damper Note 3 NS UL-555 l
Baltnce of system components are Class E or Class L Turbine Building Ventilation System (VTS)
Location: Turbine Builaing n/a Shutoff. Isolanon, and L
NS ANSI /AMCA-Balancing Dampers 500 n/a Fire Dampers Note 3 NS UL-555 Balance of system components are Class L Containment Leak Rate Test System (VUS)
Location: Auxiliary Building VUS PL-V015 Main Equipment Hatch Test B i
ASME 1112 Connection VUS-PL-V016 Maintenance Equipment B
1 ASME 111-2 Hatch Test Connection VUS-PL V017 Personnel Hatch Test B
I ASME 1112 Connection VUS-PL-VOIS Personnel Hatch Test B
1 ASME III 2 Connection VUS-PL-V019 Personnel Hatch Test B
1 ASME 111-2 Connection VUS-PL-V020 Personnel Hatch Test B
ASME III 2 Connection l
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Revision: 12.s.nimo2normon Draft,1997 3.2-76 3 Westinghouse
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FAX. to DINO SCALETTI April 16,1997 CC:
Sharon or Dino, please make copies for:
Diane Jackson Ted Quay Gordon Israelson j
Don Hutchings Bob Vijuk Brian McIntyre i
OPEN ITEM #277 (M9.4.3-3) i To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 45 calendar days away (33 business days). The relevant documentation related to Open Item #277 (M9.4.3-3) is SSAR Subsection 9.4.3 (pertinent pages are attached). This material was submitted to you in Revision 7 of the SSAR on May 6,1996 (more than 11 months ago). It is requested NRC review this material and i
provide definitive action for Westinghouse or provide direction to change the status of this item.
We recommend " Action N" or " Closed."
Jim Winters 412-374-5290 I
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AP600 Open Item Tracking Syit;ns Database: Execctiva Samanias7 Datb: 4/16/97 '
Selection:
[ item no] between 277 And 277 Sorted by Item #
Iacun DSER Section Tie-/Demiption Resp (W)
NRC-
- No.
Branch Question Type Detail Saatus Engineer Status Status 1stier No. /
Date -
277 NRR/SPLB 94.3 MTG-OI Winters /BPC Action N Action W l description in Section 9.4.3.2 of the SSAR and data in Tables 9.4.3-1 and 9.4-1.3-3 (RADIOLOGIC M9
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[ Action N _SSAR subsection 9.4.3, Revision 7, has been revised to reflect the deletion of the effluent holdup tank exhaust air llEPA filiers.
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- 9. Auxiliary Systems f
that close to isolate the auxiliary and annex buildings from the outside environment when high airborne radioactivity is detected in the exhaust air duct. The supply and exhaust ducts are configured so that two building zones may be independently isolated. The annex building, adjacent auxiliary building staging, equipment areas, and rooms served by the radiation i
chemistry laboratory ventilation subsystem are aligned to one zone. The other zone includes primarily radwaste equipment rooms, pipe chases, and adjacent access corridors located in the auxiliary building. A radiation monitor is located in the exhaust air duct from each zone.
The exhaust air fans are located in the upper radiologically controlled area ventilation system equipment room at elevation 145'-9" of the auxiliary building. The exhaust air ductwork is routed to minimize the spread of airborne contamination by directing the supply airflow from the low radiation access arers into tlieTa3foactive ~ equipment and-rooms with a greater potential for airbome radioactivity. Additionally, the exhaust air ductwork is ted to the radioactive waste drain system (WRS) sump to maintain the sump atmosphere at a neg air pressure to prevent the exfiltration of potentially contaminated air into the surrounding
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area. The exhaust air ductwork is connected to the radwaste effluent holdup tanks to prevent the potential buildup of airbome radioactivity or hydrogen gas within these tanks. De exhaust fans discharge the exhaust air into the plant vent for monitoring of offsite airborne
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radiological releases.
The ventilation airflow dilutes potential airbome contamination to maintain the concentration at the site boundary within 10 CFR 20 (Reference 21) allowable effluent concentration limits and the intemal room airbome concentrations within 10 CFR 20 occupational derived air concentration (DAC) limits during normal plant operation.
Unit coolers are located in the normal residual heat removal system (RNS) and chemical and volume control system (CVS) pump rooms because they have significat cooling loads on an intermittent basis when large equipment is operating.
Each unit cooler is sized to accommodate 100 percent of its corresponding pump cooling load. De unit coolers are provided with chilled water from redundant trains of the central chilled water system (VWS) low capacity subsystem. De normal residual heat removal pump room unit coolers have two cooling coils per unit cooler so that chilled water supplied by either train A or train B alone can support concurrent operation of both normal residual heat removal system pumps. De two chemical and volume control makeup pump room unit coolers are connected to redundant tnins of the chilled water system; however, operation of either the train A or train B unit cooler alone maintains the common makeup pump room temperature conditions and supports operation of either makeup pump.
Heating coils are located in the supply air ducts serving plant areas that require supplemental heating during periods of cold outside air temperature conditions. Electric unit heaters provide supplemental heating in the middle annulus, ne upper annulus is separated from the middle annulus area of the auxiliary building by a concrete floor section and flexible seals that connects the containment steel shell to the shield l
building. The annulus seal provides a passive barrier during normal plant operation or when i
k Revision: 11 Y W95tingh0088 9.4-29 February 28,1997
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- 9. Auxulary Systems b ('tifill l
_9.4.3.2.3 System Operation 9.4.3.2.3.1 Auxiliary / Annex Building Ventilation Subsystem Normal Plant Operation N
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1 During normal plant operation, both supply air handling units and both exhaust fans operate) f continuously to ventilate the areas served on a once-through basis. he supply is modulated to maintain the areas served at a slightly negative pressure differential with respect to the outside environment. De exhaust airis unfiltered and directed to the plant vent for discharge and monitoring of offsite gaseous releases.
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De temperature of the supply air is controlled by temperature sensors located in the supply air ducts. When the supply air temperature is low, the face and bypass dampers across the supply air hot water heating coil are modulated to heat the supply air. Local thermostats operate supply duct heating coils and unit heaters to provide supplemental heating for building areas that have conductive heat loss to the outside environment during periods of cold outside temperature conditions. When the supply air temperature is high, the flow of chilled water is modulated to cool the supply air. The ventilation air is continuously monitored by smoke monitors located in the common ductwork downstream of the supply air handling units and upstream of the exhaust fans.
A supply air handling unit is automatically shut down if one of the following conditions is detected:
Airflow rate of the fan is below a predetermined setpoint Supply air temperature is below a predetermined setpoint 1
Each chemical and volume control system makeup pump and normal residual heat removal system pump unit cooler automatically starts whenever the associated pump receives a start signal or a high room temperature signal.
j 2
De gaseous radwaste equipment areas have sufficient ventilation to remove hydrogen gas that may leak from the radwaste equipment into the equipment rooms to maintain the concentration of hydrogen below a safe level of about I percent. Instrumentation available to monitor hydrogen concentration is listed in Table t 1.3 2.
Absormal Plant Operation If high airborne radioactivity is detected in the exhaust air from the auxiliary or annex buildings, the supply and exhaust duct isolation dampers automatically close to isolate the l~
affected area from the outside environment. De containment air filtration system mitigates the exfiltration of unfiltered airbome radioactivity by maintaining the isolated zone at a slightly negative pressure with respect to the outside environment and adjacent unaffected
[(
h Revision: 11
$ Westkigh0088 9.4 33 February 28,1997 4
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1 FAX to DINO SCALETTI April 16,1997 CC:
Sharon or Dino, please make copies for:
Diane Jackson Ted Quay Don Lindgren Don Hutchings Bob Vijuk Brian McIntyre OPEN ITEM #298 (M9.4.9-1)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all j
Westinghouse submittals by May 30,1997. This is just 45 calendar days away (33 business days). The relevant documentation related to Open Item #298 (M9.4.9-1) is in Revision 11 of the SSAR, Table 3.2-3 (Sheet 58 of 64) dated February 28.1997 and is discussed further in our letter to NRC # NSD-NRC-5012 sent to you March 5,1997. Pertinent pages of these documents are attached. These were transmitted over one month ago.
It is requested NRC review this material and provide defimitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed."
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Jim Winters 412-374-5290
AP600 Open' Item Tracking Sy;t m Dat:. base: Execttiv2 Summary Date: 4/16/97 Selection:
Dtem no] between 298 And 298 Sorted by item #-
' leem DSER Section Thiption Resp (W)
NRC No.
Branch Question
. Type' Detail Status Engineer Status Staus letter No. /
Dale 298 N*A/SPLB 9.4.9 MTG4M Winners /SCS Action N Aaion W.
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,M9I4.9-I (TURBINE BUILDING VENTIl5 TION SYSTEM) Sectid919 of ESSAR d$nbes the systemNflyT IEIUhign parametirs ~
! or system components or piping, instrumentation degram, and classification of the VTS system and_w pe.s in Table 3 2-3 of the SSAll i
,ClosM - The turbine building ventilation sysicm serves no safety - related functions and has no safety design
- basis. None ofits components are classified A.B.C,or D. The system and component description provided in SSAR subsection 9.4.9, Revision 7, provides sufficient detail for a determination of the system's effect on plant safety.
Action W-Same as OITS 8 285
, Action N - Response provided by NSD-NRC-97-5012 of 3/5/97. SSAR change provided in Revision l1. jww N
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l Westin Energy Systems sex 355 Electri Corporation Pittsbutgh Pennsylvama 15230-0355 i
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NSD-NRC-97-5012 DCP/NRC076 l
Docket No.; STN-524X)3 March 5,1997 I
I Document Control Desk U. S. Nuclear Regulatory Commission
. Washington, DC 20555 L
ATTENTION: T.R. QUAY I
(
SUBJECT:
RESPONSE TO RAI 410.295 CLASSIFICATION OF MECHANICAL AND FLUID l
SYSTEMS i
Dear Mr. Quay:
Attached are responses to NRC RAI 410.29 and a related open item (OITS #289). Also attached is a copy of SSAR subsection 3.2.4 that shows the changes made in Revision 11 that implements the l
commitment in the response. Table 3.2-2 has been revised in SSAR Revision 11 to include all fluid and mechanical systems, identify the classification for all components, and provide classification of fire dampers. Table 3.2-3 has also been revised in SSAR Revision 11 to include the changes due to
~ her responses and design changes identified to date.
ot j
This response will permit Plant Systems Branch to close these items and finalize its input to FSER section 3.2.
l i
If you have any questions please contact D. A. Lindgren at (412) 374-4856.
l$4x Brian A.
clntyre, Manager Advanced P! ant Safety and Licensing jml Attachment ec:
Diane Jackson, NRC (w/ Attachment) j i-W. Huffman, NRC (w/ Attachment) i l-
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Attachment to NSD-NRC-97-5012 RAl# 410.295 (3481)- NRC Letter 8/15/1996, SSAR Table 3.2-3, AP600 Classification of Mechani-cal and Fluid Systems, Components, and Equipment:
Westinghouse needs to revise Table 3.2-3 to provide the classification of the following fluid a.
systems and their associated generalized equipment:
1.
Radiologically Controlled Area Ventilation System (VAS)
- 2. Containment Recirculation Cooling System (VCS)
- 3. Health Physics and Hot Machine Shop HVAC syste S) adioactive Waste BMgMSy:& (V4,
- 5. Turbine Building Ventilation System (VTS)
- 6. Annex / Auxiliary Nonradioactive Ventilation System (VXS
- 7. Liauid Waste Management sycem
- 8. Gaseous Waste Management System
- 9. Radiation Monitoring System
- 10. Main Steam System
- 11. Condensate Storage System
)
- 12. Reactor Coolant Pressure Boundary (RPCB) Leakage Detection and Monitoring System b.
In the previous version, there was a " Location" column in the table, which is useful to the reviewer. It was removed from the table in Revision 8. Bring the location information back to the table.
Westinohnuse Remonse
/
a.
Table 3.2-3 focuses on the classification of safety related (Class A, B, or C) components and
'f equipment in mechanical and fluid systems. Items that are AP600 equipment Class D or
)
equivalent are identified in a general basis. The table is being revised to include the fluid and 2
mechanical systems in AP600. Items that are not Class A, B, C, or D are not individually identified. Systems that are electrical or instrumentation systems are not included in this table.
This is consistent with the guidance in Regulatory Guide 1.70. De components in the incore instrumentation sy: tem that have a pressure boundary function are included in the table.
Responses for the specific systems follows. The text in subsection 3.2.4 referencing the contents of the table has also been revised.
- 1. The radiologrally controlled area ventilation system (VAS) is included in the table. The Class D toom coolers and valves that provide a Class D function are identified. The balance of the equipment in the VAS is Class E.
- 2. Le equipment in the containment recirculation cooling system (VCS) is Class E or Class L.
- 3. The equipment in the health physics and hot machine shop HVAC system (VHS) is Class E.
- 4. The equipteent in the radioactive waste building HVAC System (VRS) is Class E, Class L, Class F, or Class R.
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Attachment to NSD-NRC-97-5012 l-i s/
N l
S.
The equipment in the turbine building ventilation system (VTS) is Class E, Class L, or j
}
Class F.
6.
The annex / auxiliary nonradioactive ventilation system (VXS) is included in the table. The air handling units and dampers that provide a Class D function are identified. The balance of the equipment in the VAS is Class E.
7.
The liquid radwaste system (WLS) is included in Table 3.2-3.
8.
The gaseous radwaste system (WGS) is included in Table 3.2-3.
9.
The radiation monitoring system (RMS) is not a fluid system and is not included in Table 3.2-3. The system is discussed in Section 11.5. Most radiation detectors are included as a part of the system they monitor. Table i1.5-1 identifies the safety-related monitors.
l Subsection 7.1.4 provides information on qualification and other requirements for safety-l related monitors.
- 10. The main steam system (MSS) is the nonsafety-related portion of the main steam line and l
associated piping. Equipment in the MSS is Class E. The safety related portion of the main steam line is included in the steam generator system (SGS).
i
- 11. The condensate system (CDS) is not safety-related. Equipment in the CDS is Class E.
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- 12. The reactor coolant leak detection function is provided by components and subsystems in a I
number of systems, it does not rely on a dedicated leak detection system. Subsection 5.2.5 addresses the leak detection approach and identifies the systems which include the equipment required for leak detection,
- b. The location information in the previous revision of the table merely identified in which building a component was located. Safety-related equipment is either in the containment or auxiliary building. The location for all or most of the equipment will be identified on the line with the system name. Systems such as the plant gas system that have components in several of the buildings are idemified as being located in "Various" buildings. More specific informa-tion on the locations of safety-related equipment is provided in Table 3.11-1 and the associated information in Table 3D.5-1. Location by fire area in provided in Appendix 9A.
OITS 281 and NRC letter dated October 17,1996, item 7. e. (2) (OITS 289) identified the request to include' fire dampers in. Table 3.2-3.
Wecinohnute Reenanam I
Fire dampers have been added to the table in SSAR Revision 11. The approach used is simdar to the l
approach used for Class D equipment.
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4 Attachment to NSD-NRC-97-5012 i
i l.
3.2.4 Application of AP600 Safety-Related Equipment and Seismic Classification System l
The application of the AP600 equipment and seismic classification system to AP600 systems and components is shown in Table 3.2-3. Table 3.2-3 lists safety-related and seismic category I mechanical and fluid system component and associated equipment class and seismic category as well as other related information. The table also provides information on the systems that I
contain Class D components. Additional information on the Class D functions of the various I
systems can be found in the description in the SSAR for the systems. Mechanical and fluid I
systems that contain no safety-related or Class D systems are included in the table and general
. l information provided on the system. Supports for piping and components have the same l
classification as the component or piping supported. Suppons for AP600 equipment Class A, B, and C mechanical components and piping are constructed to ASME Code, Section Ill, Subsection NF requirements. The principle construction code for supports for nonsafety-related components and piping is the same as that for the supported component or piping.
I Following the name of each system is the building location of the system components. Some of I
the systems supply all or most of the buildings. This is indicated by identifying the location as I
various. Where a system includes piping or ducts that only passed through a building without I
including any components that building is generally not included in the list.
ym The % wine N Ed-L me systems in Table 3.2-3. The three letters in the beginning of each
/
line is the acronym for the system. The systems included in Table 3.2-3 are listed alphabetically I
by three letter acronym. The ;;;'= pF: d !: " "-- b; = =2;=:r i-C'--- ^., B, C,
D :d b; = S Of==p:
- =h9d in 6: d':. Those systems marked with an 1
I asterisk
- are electrical or instrumentation systems and are not included in Table 3.2-3. The I
components in the incore instrumentation system that have a pressure boundary function are l
included in the table. See Section 3.11 for identification of safety-related electrical and instrumentation equipment.
NSSS/ Steam Generator Controls And Auxiliaries 1
BDS Steam Generator Blowdown System CNS Containment System CVS Chemical and Volume Control System PCS Passive Containment Cooling System PXS Passive Core Cooling System RCS
- Reactor Coolant System RNS Normal Residual Heat Removal System RXS Reactor System SGS-Steam Generator System L
Nuclear Control and Monitoring i
- DAS Diverse Actuation System l'
- IIS Incore Instrumentation System L
- OCS Operation and Control Centers l
- PMS Protection and Safety Monitoring System PSS Primary Sampling System i
- RMS Radiation Monito ing System b
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Attachment to NSD-NRC 97-5012 VHS Health Physics and Hot Machine Shop HVAC System VMontauunent HTdrogsu Cuend c"
'AS Radwaste Building HVAC System VTS Turbine Building Ventilation System VUS Containment Leak Rate Test System uuu.2 unucu n.m aimiu VXS Annex / Auxiliary Nonradioactive Ventilation System VYS Hot Water Heating System' VZS Diesel Generator Building Ventilation System Turbine-Generator Controls and Auxiliary CMS Condenser Air Removal System HCS Generator Hydrogen and CO Systems HSS Hydrogen Seal Oil System LOS Main Turbine and Generator Lube Oil System
- TOS Main Turbine Control and Diagnostics System Material Handling FHS Fuel Handling and Refueling System MHS Mechanical Handling System Piping Services CAS Compressed and Instrument Air Systems DOS Standby Diesel and Auxiliary Boiler Fuel Oil System FPS Fire Protection System i
PGS Plant Gas Systems i
PWS Potable Water System Non-Class IE Power Systems
- ECS Main AC Power System i
- EFS Communication Systems
- EGS Grounding and Lightning Protection System
- EHS Special Process Heat Tracing System
- ELS Plant Lighting System
- EQS Cathodic Protection System Non-Nuclear Controls and Monitoring
- DDS Data Display and Pacessing System I
- MES Meteorological and Envi=ranental Monitoring System l
- PLS Plant Control System
- SES Plant Security System SSS Secondary Sampling System
- TVS Closed Circuit TV System 5
m
I I
- 3. Design of Structures, Components, Equipment, and Systems i,
n-I i
l Table 3.2-3 (Sheet 58 of 64)
AP600 CLASSIFICATION OF MECHANICAL AND FLUID SYSTEMS, COMPONENTS, AND EQUIPMENT Tag Number Description AP600 Seismic Principal Con. Comments Class Category struction Code i
Containment Hydrogen Control System (VLS)
Location: Containment n/a Hydrogen Igniters D
NS Manufacturer Provides Std.
Hydrogen Control Following Severe Accidents VLS-MY-E01 A Catalytic Hydrogen C
I Manufacturer Recombiner A Std.
VLS-MY-E01B.. - Catalydd Hfdf6 gen
~ C' mufacturer f'~
Recombiner B Std.
a Fire Dampers Note 3 NS UL-555
,-l Balance of system camponents are Class E or Class L l
Turbine Building Ventilation System (VTS) location-Turbine Building I
n/a Shutoff, Isci:!bn, and L
NS ANSI /AMCA-l Balaneirg Dampers 500 l
n/a Fire Dampers Note 3 NS UL-555
/
l Balance of system components are Class L
/
Conininment Leak Rate Test System (VUS) on: Auxiliary Building b~-V015-Cc.:.=;& :
Main Equipment Hatch Test B ASME m-2 VUS-PL-V016 Maintenance Equipment B
1 ASME DI-2 Hatch Test Connection VUS-PL-V017 Personnel Hatch Test B
I AShE m-2 Connection VUS-PL-V018 Personnel Hatch Test B
I AShE m-2 Connection VUS-PL-V019 Personnel Hatch Test B
1 ASSE m-2 Connection VUS-PL-V020 Personnel Hatch Test B
I AShE m-2 Connection VUS PL-V021 Personnel Hatch Test B
1 ASME m2 Connection VUS-PL V022 Personnel Hatch Test B
1 ASME m-2 Connection VUS-PL-V023 Fuel Transfer Tube Test B
I ASME W-2 Connection Revision: 11 o
February 28,1997 Ng 3 Westingh0use 3.2 76
l FAX to DINO SCALETTI l
April 16,1997 CC:
Sharon or Dino, please make copies for:
Ted Quay Don Hutchings Bob Vijuk Brian McIntyre OPEN ITEM #300 (M9.4.10-2)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 45 calendar days away (33 business days). The relevant documentation related to Open item #300 (M9.4.10-2)is in my fax to Diane Jackson / Tom Kenyon sent January 9,1997 which transmi ed a markup of SSAR Subsection tt 9.4.10.2.1.1. These proposed changes to Subsection 9.4.10.2.1.1 were incorporated into Revision 11 of the SSAR dated February 11,1997. Pertinent pages of these documents are attached, it 1
is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed."
s Jim Winters 412-374-5290 l
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' AP600 Open It:mi Tracking Systema Detr. base: Emeettiva Susmaanry
- Dathil6/97 Seleef m:
[ item nol between 300 Ami 300 Sorted by item #
~
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USER Secten Title / Description Resp
-(W)'
NRC-Detail 54asus Engineer status -
Status No.
' Branch Question Type--
~ -.
- - _ - -._. Wr No.! I-Dec
- 300I NRR/SPt.B -
9.4.10 MTG4)I
. Winters /BRC Action N Action W
[M9Il0-2 (DIESEL' GENERATOR BUILDING llWAC SY' STEM) To deschine the cit Aof con 55ivinance widi GDCi7[a's'it ielsees EeNd
~
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! proper functioning of the standby onsiac ac electric power system, Westmghouse should confum that the VZS standby exh are equipped with the air fihers and that the louver locations for outside air intake conform with the guidance of NURECKR-0660 [6.1 m (20 ft) he grade to control the dust and oeher particulates. Revise Figwe 10A.10-1 of the SSAR accordingly.
]
[Closi[-~As iniscated in the "AP600 O_
,. --* staaement for GDC 17 in SS R subsecame 3.1.l, Revision'7,
[
~
ohe diesel generasors are not part of the safety-selated power sources for AP600. As such NURECKR-0660 j
need not apply to the design of the diesel generators, the diesel generator building or the dicsci generasor r
, building IIVAC. However, since NURECKR-0660 does provide prudent design gaulance, the diesel generasor building heating and ventilation system does not comply as follows: The air intake to the senice module, which houses the major electncal equipment associated with the DG unit, is located as high up on the building
.[
,as practical jOutside air to the service module is filtered to prevent the spread of dust and dirt onto electrical equipment.
L jln all operating inodes, air supplied to electncal equipment areas is filtered. -
'Electncal eqmpment cabinets are specified to be dust tight.
l Air pmvided to the engine scent is filtered and is taken through intakes locased as high as in the building t
N as practical i
b Coinbustion air is taken from the outdoors via a combustion air cleaner, not from the engine room.
g llie stanJby exhaust ventilation subsystem operales only when the diesels are running and supplies l
\\
ventilation only to die engine rooms. The supply of filtered air to electncal equipment is unaffected by I
operation of the standby exhaust ventilation subsystem.
i
'i Action W - Expand section 9A.10.2.1.1 to descnbe air intakes as being as high in the building as possible.
f Anction N - FAX provided with SSAR markup on January 9,1997.-
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Westingh00S8 FAX COVER SHEET i
1 RECIPlENT INFORMATION SENDER INFORMATION OATE:
JaauW 9, /99 7 NAME:
L Q,ygc TO:
LOCATION:
ENERGY CENTER -
Dinae Incresoefxm t<vnAa EAST PHONE:
FACSIMILE:
PHONE:
Office: t/r 2. - 3 7 '/ - s t 90 COMPANY:
Facsimite:
win:
284 4887 d i A/AC-outside: (412)374 4887 LOCATION:
i Cover + Pages 1+/
The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please cati:
WIN: 284 5125 (Janice) or Outside: (412}374 5125.
COMMENTS:
Dave
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- 9. Auxiliary Systims The system maintains the following room temperatures based on ambient outside air temperature conditions of 95'F (summer) and -5'F (winter):
Design Temperature Area Minimum Maximum
('F)
("F)
Diesel Generator Area Diesel Generator On...
. None 130 Diesel Generator Off
......50 105 Service Module Diesel Generator On 50 105 Diesel Generator Off...
50 105 Diesel Oil Transfer Module Enclosure
. 50 105 9.4.10.2
System Description
The diesel generator building heating and ventilation system is shown in Figure 9.4.10-1.
De system consists of the following subsystems:
q Normal heating and ventilation subsystem Standby exhaust ventilation subsystem I
Fuel oil day tank vault exhaust subsystem Diesel oil transfer module enclosures ventilation and heating subsystem
).4.10.2.1 General Description
{
9.4.10.2.1.1 Norma. IIcating and Ventilation System i
The normal heating and ventilation subsystem serves the diesel generator building. Each diesel generator train is provided with independent ventilation and heating equipment for the building areas serving that diesel generator train.
Each normal heating and ventilation subsystem for a diesel generator train consists of one 100 percent capacity engine room air handling unit which ventilates the diesel generator room, one 100 percent capacity service module air handling unit which ventilates the electri.al equipment service rnodule, an exhaust system for the fuel oil storage vault and electric unit heaters in thg diesel generator area.S.< sam' /ouvs 4 4.c4. un'f2 A" loc a.*d " 8 4'j l 'a
- dum! pas owtely est as possou<.
The engine room air handling units are located above the electrical equipment service module with supply and retum ducts in the diesel generator room.
The service module air handling units are located above the service module with supply and return ducts into the module.
h-Revision: 10 December 20,1996 9.4-60 W Westingh0use
i
- p _=:
- 9. Auxiliary Systems n
Service Module Diese! Generator On.
. 50 105 Diesel Generator Off.....
. 50 105 l
Diesel Oil Transfer Module Enclosure...
50..
105 j
9.4.10.2
System Description
ne diesel geverator building heating and ventilation system is shown in Figure 9.4.10-1.
l The system consists of the following subsystems:
Normal heating and ventilation subsystem Standby exhaust ventilation subsystem Fuel oil day tank vault exhaust subsystem Diesel oil transfer module enclosures ventilation and heating subsystem j
9.4.10.2.1 General Description 9.4.10.2.1.1 Norral Heating and Ventilation System ne nonnal heating and ventilation subsystem serves the diesel generator building. Each diesel generator train is provided with independent ventilation and heating equipment for the building areas serving that diesel generator train.
(
Each normal heating and ventilation subsystem for a diesel generator train consists of one 100 percent capacity engine room air handling unit which ventilates the diesel generator room, one 100 percent capacity ** m^0& au imm!!m3 mui olucl. a..:^*" the electrical equipment i
se u e, an exhaust system for the fuel oil storage vault and electric unit heaIE!Tm4be I
diesel generent area. Air intake louvers for these units are located as high in the 1
I generator building wall as possible.
He engmc mum air-handling-units am ic~dd3b5Ve the electrical equipment service module with supply and return ducts in the diesel generator room.
The service module air handling units are located above the service module with supply and return ducts into the module.
Electric unit beaters are provided in the diesel generator room to maintain the space at a minimum temperature of 50*F when the diesel generators are off.
9.4.10.2.1.2 Standby Exhaust Ventilation Subsystem ne standby exhaust ventilation subsystem for each diesel generator room consists of two 50 percent capacity roof mounted exhaust fans and motor operated air intake dampers mounted in the exterior walls of the room.
l
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Revision: 11
_h Febatary 28,1997 9.4-62 W Westinghouse
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FAX to DINO SCALETTI April 16,1997 CC:
Sharon or Dino, please make copies for:
Diane Jackson Ted Quay Don Hutchings Bob Vijuk Brian McIntyre OPEN ITEM #304 (M9.4.11-3)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 45 calendar days away (33 business days). The relevant documentation related to Open item #304 (M9.4.11-3) is in my fax to Diane Jackson sent January 28,1997 and in SSAR Figure 9.4.4.7-1. Pertinent pages of these documents are attached.
it is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this item. As shown on the attached, we now recommend " Action N" or " Closed."
Jim Winters 412-374-5290 l
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o.
AP600 Open item Tracking System Database: Executive Samma 7 Date: 4/16/97 Selectian:
[ item no] between 304 And 304 Sorted by item O ltem DSER Section Titic/Descriptum Resp (W)
NRC No.
Branch Question Type Detail Status Engineer Status Status le.tter N. o. /. _... Date 304 NRR/SP1 B 9 4 II MIG 4)I Winters /BRC Action N Action W
[M9 4 Il-3 (IIEALTil PilYSICS AND IIOT MACilINE SilOP lIVAC SYSTEM) Clanfy the' number'of rsfiation nmiters with associated MCR
~
,high and high-high alarms and number of filtration unit for the hot machine shap ait provided and reiise affected SSAR section, figure, table, and
'"E._
- E I-I l Closed - Dere is a single radianEm rnonitEin the exhaus't of the healthiphyEs a~ndInd madtine sh[.. _
~
~~
~
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IllVAC system to the plant vent stact This is indicated in SSAR subsection 9 411.2.I, Revision 7 and I
!in Subsection i1.5, Revision 6.
Action W - Provide justification for deletkm ofIIEPA filter and desenbe results in SSAR.
Aguy N - Jusification provided in FAX on I/28/97, no change to SSAR required.__
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FAX to DIANE JACKSON 1
January 28,1997
- This is in response to item 7.i.(1) of your 10/17/% letter, item 7.i.(1) of our 12/17/96 telecon and I
'OITS item 304. HEPA filtration is not required on the VHS exhaust to the plant vent because the j
high radiation alarm in the exhaust line would lead the operators to terminate discharge on high radiation. This approach is consistent with the VRS and the two VAS exhaust paths to the plant vent shown on Figure 9.4.7-1 (Sheet 1 of 2). We recommend that the "NRC Status" for item 304 be changed to " Action Y
l I
Thanks I
Jim Winters
]
412 374-5290 o
i cc:
McIntyre Cummins Hutchings 1
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- 9. A xiH:ry Systems 7
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"%Ws Figure 9.4.7-1 (Sheet I of 2) l Containment Air Filtration System Piping and Instrumentation Diagram Revision: 7 Westinghouse April 30,1996 ff 9.4-111
WESTINGHOUSE CALCULATION SHEET TITLE PAcg PROJECT AUTHOM DATE CH K'D. S V DATE VERIFIED BY DATE 50, CALC. NO.
FILE NO.
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DATE WESTINGHOUSE FORM 66213H (3/091 l
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l' Westinghouse Response to Staff Comments on l
Comparison Between the Adverse System Interaction Report and Emergency Response Guidelines 1
Completeness and Accuracy Some issues were noted with regard to the completeness and accuracy of the ASI report.
Specific examples follow:
(1)
Not all of the SSCs noted in Section 2.1 appear in Table 2.1.
- pH adjustment of containment recirculation does not appear on the top of the Table.
- RCS pressure control (pressurizer spray), reactor vessel head vent, and the steam generator system are missing from the initiating system column of the Table.
Westinghouse Response:
The ASI report has been updated to address this comment. The left hand column of Table 2-I has been modified such that pressurizer heaters has been relabeled RCS pressure control.
Section 2.2.2 has been modified to address pressurizer spray interactions.
The steam generator system was not added to Table 2-1 as a seperate line item because its interactions are accounted for in the sections on MFW pumps, SFW pumps, and the SG blowdown system. The pH adjustment function was not added to the top row of Table 2-1 because this function is now performed by baskets located inside containment containing trisodium phosphate (TSP). No system interactions with the TSP baskets have been identified.
- W (2)
In Section 3.1 a statement is made that "the ER6tFRGs are symptom-based, thus they do not allow the operator (to) make knowledge-based decisions." This is not in accordance with the current philosophy of ERGS and EOPs in the industry. In particular, one of the documents (NUREG/CR-6208) authored by and provided by WEC for the Operating Experience Review of Chapter 18, makes a strong case for operator knowledge-based reasoning (particularly situation assessment and response planning) during the execution of emergency procedures. The statement in the ASI report should be revised or removed.
This section also states that when a red path is reached in the status trees, the operators go to the FRGs. This is also true for orange paths but is not noted.
d I
=
Westinghouse Response:
ff45 The ASI report has been modined. The statement reads that the "the E-RGTRG are symptom-based, thus they do not allow the operator to make knowledge-based decisions contrary :: 'he rule-based decision philosophy of the gun. nu,. This is consistent with NUREG/CR-6208.
W5 In addition, the comment regarding the response to an " orange path" within the ERGS has been incorporated.
(3)
Section 3.2 states that each adverse system interaction identified in Section 2.0 is examined from the point of view of human errors of commission. However, those adverse interactions of Section 2.2.14 are not included here.
Westinghouse Resp (mse:
Section 3.2, Table 31 has been revised to address this comment.
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(4)
Section 3.3, Commission Error Related to Economic Consequences, concludes that such an error is "not credible" without seriously considering the possibilities and explaining why it may not be credible. The only reason given is that procedures require it. However, there have been many cases where procedures have not been followed and there are also cases where operators have been reluctant to take actions that have economic consequ: aces (e.g. Davis Besse). The ASI report should be revised to acknowledge that such an error is possible but of low probability. It may be appropriate to discuss if the PRA accounts for such an occurrence.
Westinghouse Response:
Section 3.3 has been revised to address this comment as follows:
3.3 Commission Error Related to Economic Consequences Another set of potential commission errors is related to economic consequences. In some events sequences, the operators may be reluctant to actuate safety-related system if it may have a large adverse economic consequences to the plant. An example of such an action often cited is the reluctance to start emergency feed and bleed operations in a conventional plant. In current plants, emergency feed and bleed operation can be required to mitigate loss of heat sink events, where the operator needs to cool the core without the use of the steam generators. To accomplish this, the operators " feed" with the high head safety injection pumps, and " bleed" with the PORVs connected to the pressurizer. However, the operators are reluctant to perform this operation since operation of the PORVs would lead to a cleanup p
of containment; therefore, this operation is used as a last resort. A potential failure is the d
i l
operators may continue to try to recover secondary cooling, rather than go to feed and bleed until it is too late.
In the AP600, a counterpart to feed and bleed operation is related to actuation of the automatic depressurization system. The ADS valves, in conjunction with the CMTs, accumulators, and IRWST provide automatic emergency core cooling in response to loss of coolant accidents. However, the important differences between manual actuation of ADS in the AP600 and the use of feed and bleed cooling in current plants are:
- the ques and man-machine interfaces available for manual ADS actuation
- the consequences of manual ADS actuation These differences are explained below.
The decision making process that the operator is faced with in the AP600 is much different than that of current plants. To establish feed and bleed cooling in current plants, the operator must make a decision to initiate feed and bleed soley based on conditions in the plant.
Therefore, as the RCS conditions slowly degrade due to the loss of the RCS heat sink, the operator may continue to try to re-establish the normal, front-line safety system (i.e. steam generator cooling) and forego the onset of feed and bleed cooling.
In the AP600, ADS is an automatic safety feature that is used in response to design basis accidents. It is not a "last resort", but rather a front-line safety system that would be expected to actuate for loss of-coolant accidents. There are two scenarios where manual actuation of the ADS is required. These are:
- following failure of automatic actuation of ADS on low CMT water level
- following a loss of all RCS heat sinks (SGs and PRHR)
The use of ADS following a failure of automatic actuation of ADS on low CMT water level is not comparable to manual actuation of feed and bleed cooling in current plants. In this case, the automatic protection system should have automatically sctuated ADS, and therefore the operator has a clear and unambiguous que.that manual ADS actuation is required.
The use of manual ADS following a loss of all RCS heat sinks is more similar to feed and bleed cooling in current plants. In current plants, the use of feed and bleed cooling is used in beyond-design basis events, where multiple failures result in a loss of all RCS cooling via the steam generators, in the AP600, the use of manual ADS to provide feed and bleed cooling is even more unlikely. The failure of an RCS heat sink occurs only after the failure of both the normal and startup feedwater systems, and requires the complete failure of the safety-related PRHR heat exchanger. It is only after the failure of the PRHR, that the operators would then use feed and bleed via manual actuation of ADS.
In this scenario, the emergency procedures would have provided clear and unambiguous guidance to actuate ADS manually. In addition, the presentation of the emergency operating procedures (EOPs) via computerized procedures will provide the operator with clearer
guidance than provided in todays plants. Due to the importance of ADS, the presentation of the information necessary to que the operator to manually actuate ADS will be given a sufficiently high priority in the AP600 MMIS design process.
The other major difference between current plants and the AP600 is that the AP600 is designed to minimize the consequences of ADS operation. In current plants, even limited feed and bleed can result in rupturing of the pressurizer relief tank and causing reactor j
coolant to be spilled into containment. In the AP600, the first three stages of ADS connected to the pressurizer discharge to spargers located in the IRWST. This minimizes the
{
consequences of the first three stages of ADS operation to the containment environment. As i
discussed in section 2.0 of this report, following a ADS without a LOCA, if the normal residual heat removal pumps were to operate, the CMTs will not drain to the fourth stage ADS setpoint, and significant containment flooding will not occur. Even in the event of a full ADS (i.e. fourth stage ADS opening), the RCS loop compartments would fill with water, but significant flooding elsewhere in containment would be avoided.
In summary, failure of the operators to manually actuate ADS is a very unlikely event.
Operator acts of commission to prevent ADS are not credible. Failure of the operators to manually actuate ADS would require the operators to ignore and / or override emergency procedures. Wheras in current plants, the operators faliure to actuate feed and bleed has been noted, it is much less likely in the AP600 due to the fundamental design difference between automatic depressurization, and manual feed and bleed cooling. The use of advanced MMIS and computerized procedures, as well as the clear and unambiguous guidance provided in the ERGS provides assurance that the operator will not be tempted to overrides system actuation of ADS when required. Furthermore, the economic consequences of ADS in the AP600 is much less severe than feed and bleed in current plants, and it will not pose an undue temptation to the operators to avoid its use. Therefore errors of commission related to ecomic consequences have been sufficiently considered in the design of the AP600 and are not considered credible.
(5)
Section 3.4 states there are six human actions impacting adverse system interactions, but then lists seven. Also the section does not give the criteria for selecting these seven from the larger number in Table 3.1.
Westinghouse Response:
The ASI report has been updated to provide the criteria for selecting the interactions that require further discussion in this section.
(6)
The Table 3.1 analysis of RCP interactions (Section 2.2.1) notes that no opportunity l
for human error (HE) exists and that there is no LOCA-related procedure where the l
operators are instructed to restart RCPs if CMTs are required. This response misses an important point. There are ERGS that direct the restart of RCPs. In the TMI ac-cident, the operators didnt know there was a LOCA and were not in a LOCA-related t
I
l procedure. The ASI report should address the design features that prevent restart of the RCPs if the CMTs are actuated.
Westinghouse Response:
The ASI report has been updated to address this comment. The ERGS address this interaction by providing clear and unambiguous guidance to avoid this interaction. In addition, the PMS prevents the restart of the RCPs in the presence of a valid CMT actuation signal. The following section has been added to section 3.5 (formerly 3.4).
2.2.1 - RCP - Core This adverse interaction is addressed in the design of the AP600. The automatic reactor coolant pump trip significantly reduces the potential for this interaction when compared to current plants. Inadvertent pump restart is prevented by the interlocks provided in the protect and safety monitoring system. Restart of the pumps requires the operators to manually block the safeguards actuation signal, and restore RCS conditions (i.e. pressurizer level) to pre-CMT actuation conditions. Subsequently, if pressurizer level can not be maintained, then the CMT actuation signal will be re-established on low pressurizer level, and the RCPs again will be tripped to allow proper CMT operation. Explicit reactor coolant pump restart criteria are provided in the ERGS that allow the operator to block the CMTs and restart the RCPs.
Illtopsistency with the ERGS (7)
Section 2.2.1 Reactor Coolant Pump Interactions: A potential adverse interaction noted here is similar to TMI operation of RCPs can mask the severity of an event by misguiding operators that sufficient coolant inventory is present. This section states that, "as presented in the AP600 ERGS, restart of the RCPs requires core subcooling and pressurizer level to be indicated." This is true for ERGS AES-1.2, AE-3, and AFR-l.3, but it is nol true for ERGS AES-0.1, step 10, AES-0.2, step 1, or AES-1.1, step 17.
Westinghouse Response:
The ERGS are consistent with the ASI report. As noted by the NRC, ERG AES-1.2, AE-3, and AFR-l.3 include the aforementioned CMT termination criteria prior to RCP re-start (followed by CMT isolation). The NRC correctly points out that no such criteria is included within ERG AES-0.1 (step 10), AES-0.2 (step 1), and AES-1.1 (step 17). The reason that this criteria is not included within the body of these guidelines is that entry into these guidelines is predicated on meeting the CMT termination criteria. A more detailed description is provided below:
i AES-0.1
-.. - - -. - ~ ~.
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. This ERG is entered after a reactor trip without a safeguards actuation signal.
Therefore, since a safeguards signal is not actuated, the CMT discharge valves, if 1
opened on low pressurizer water level, can be re-closed if pressurizer level is
~
recovered (Step 6).. For this case, the RCPs can then be restarted. The ERG 1
t
- background document will be revised to incorporate this insight, and state that j
detailed pump restart criteria should include precautions that pressurizer level be stable'or increasing.
1 AES-0.2 Similarly, this ERG is entered from AES-0.1, when it is determined that a natural
~ irculation cooldown is desired. No safeguards actuation signal has been generated, c
restarting the RCPs is allowed. The ERG background document will be revised to-
. incorporate this insight, and state that detailed pump restart criteria should include precautions that pressurizer level be stable or increasing.
AES-l.1 This ERG - Passive Safety System Termination, is used to recover the plant after j
actuation of the passive safety systems. It is entered from AE-0 -- Reactor Trip or
- i Safety injection, or AE-1 -- Loss of Reactor or Secondary Coolant, when the specific CMT termination criteria have been met. Therefore, the termination criteria is not repeated AES-1.1,
-(8)
Section 2.2.2 Post-Accident Interactions involving the Pressurizer (PZR) Heaters:
This section states that AP600 addresses the potential adverse interactions during an SGTR by the tripping of the PZR heaters on an S-signal actuation. However, AE-3, for an SGTR, step 11 directs the operator tu turn off the PZR heaters and does not mention at all the automatic trip.
Westinghouse Response:
The ERG is written to cover both design basis type events (where isolation of the pressurizer heaters is required) and multiple-failure events where the operator may have turned on the i
heaters to proceed with a cooldown (due to a loss of event) and subsequently a steam generator tube ruptures. Westinghouse recommends no change to the ERGS. The ERG background document will be revised to include the above discussion, and will also state.that j
step 11 is intended for the operators to block the automatic control system signal that starts the pressurizer heaters.
l
.(9)
Section 2.2.3 Chemical and Volume Control System (CVS) Makeup Pump Interactions: discusses three actions to prevent PZR overfill, including stopping the makeup pumps. AFR-1.1, Response to High Pressurizer Level, does not include any
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l of these three steps and in fact step 2 starts a makeup pump. Other ERGS (e.g. AE-0, step 17, and AFR-H.1, step 14) do contain a step to stop makeup pumps on a high PZR level. Table 3.1 states that procedures instruct the operator to trip CVS pumps, however not all appropriate ERGS appear to do so.
Westinghouse Response:
The ERGS address stopping the makeup pumps on high pressurizer level (AE-0, Step 17b; AFR-H.1, Step 14b). AFR-l.1 will be revised to more appropriately prioritize starting the Reactor Coolant Pumps to initiate letdown, and as a backup, instructs the operator to start the makeup pumps to initiate letdown. Either RCP operation or makeup pump operation is necessary to establish letdown via the normal letdown line.
(10)
Section 2.2.3 also states that during SGTR events to prevent adverse interactions, the CVS makeup line receives a signal to isolate on high SG water level. However, AE.
3, for an SGTR, does not mention this isolation. Further, step 4.b says to stop the makeup pumps, and step 4.c says to operate makeup and letdown to maintain PZR level between L(05) and L(06). Table 3.1 mentions an automatic CVS pump trip on high SG level, which is not mentioned in Section 2.2.3. Thus, the appears to be that on an SGTR with a high SG level, the automatic system will be tripping the pumps and/or isolating the makeup line, while the ERG is telling the operators to run the pump and maintain level. Table 3.1 recognizes the con 0ict here, but does not appear to think it a problem and still classifies this adverse interaction as a (1) or "no credible concern for human commission error." This issue /concem is not mentioned in the background material for ERG AE-3.
Westinghouse Response:
The ERGS will be revised to instruct the operators to verify that the makeup pumps are i
stopped on high steam generator level, i
(11)
Section 2.2.5, Makeup Control System: This section discusses boron dilution scenarios and notes that the clean water (demin) isolation valves receive auto isolation signals on reactor trip, safeguards actuation, etc. AES-0, Reactor Trip or Safety injection, and AES-0.1, Reactor Trip Response, do not mention or verify this isolation. This line is left out of Table 3.1 in an apparent oversight, and hence this interaction was not analyzed in Section 3.
Westinghouse Response:
l The ERGS (AE-0 and AES-0.1) will be revised to incorporate this comment. Table 3.1 in the i
ASI report has been revised as appropriate.
l
(12)
Section 2.2.6, Letdown Line Isolations: This section discusses letdown line isolations to prevent loss of reactor coolant outside of the reactor coolant pressure boundary and outside of containment. Of particular note is the case during shutdown operations when draining water from the RCS via letdown. Here the letdown isolation on low i
PZR level must be blocked. The section notes that in this case there is also an i
isolation on low hot leg level. ERG SDG-1, Response to Loss of RCS Inventory l
During Shutdown, step 2 states, " Verify RCS Drain Path Isolation." Neither this step nor the rest of the ERG says anything of the PZR low level isolation being blocked.
Table 3.1 incorrectly notes that LOCA-related procedures instruct the operator to open the letdown line. Also the Table does not acknowledge or analyze the potential for I
shutdown draining operations errers that have occurred in many operating plants.
Westinghouse Response The Shutdown ERGS are correct as written. The Shutdown Safety Status Tree uses low pressurizer water level and low hot leg water level as the appropriate ques to enter Shutdown Guideline SDG-1. It addresses whether the low pressurizer level signal is blocked by the block "CMT Actuation Signal Blocked." No change to the ERGS are necessary. Table 3.1
)
has been modified to state that LOCA related procedures instruct the operators to close the letdown line, as appropriate.
(13)
One particular potential adverse interaction was noted in the ERGS, that was not contained in the WCAP. Instrumentation can interact with various systems in adverse ways, particularly when there are failures or inaccuracies in the instrumentation. This type of interaction can also contribute to the possibility of human errors of commission. An example noted in ERG AE-1 was in steps 5 and 6, where improper instrument readings by the operators could result in incorrect termination of the Passive Safety Systems. A discussion of the potential for this type of interaction 4
would seem to be appropriate in the ASI report.
Westinghouse Response:
The ASI report has been updated to address this comment. The following section has been added to the ASI report.
3.4 Commission Error Related to Economic Consequences Another set of potential commission errors is related to instrumentation errors. The issue is whether the AP600 design has accounted for the potential of errors in instrumentation that cause the operators to make an error of commission. This has been addressed in two ways.
The ERGS have addressed potential instrumentation errors due to adverse environments. This has been addressed by providing instrument setpoints for operator actions for conditions when an adverse environment exists, in addition to the normal setpoints for when
!~.
I environmental conditions are normal. In addition, the AP600 advanced control room and MMIS, with computerized procedures, will assist the operator in determining when to utilize the setpoints for adverse environrnental conditions.
Another excercise that Westinghouse has perfromed to address the potential for errors in instrumentation to cause operator errors is the description of the analysis of the AP600 safety 0related instrumentation requirements provided in SSAR section 7.5. This section presents the results of this analysis, conducted in accordance with Regulatory Guide 1.97 to assure that the AP600 safety-related instrumentation has sufficient redundancy and diversity of qualified instruments to provide the operators with clear and unambiguous information following an accident. In this manner, the potential for operator errors due to failures in instrumentation has been addressed.
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Westinghouse Response to Additional Comments on Comparison Between the Adverse System Interaction Report and Emergency Response Guidelines (1)
Table 2.1 of the ASI report for IRWST interactions does not cross-reference Sections 2.3.3.1 and 2.3.3.2 under the columns for Recirculation and PRHR (for the IRWST).
Westinghouse Response:
The ASI report has been modified to incorporate this comment.
(2)
Section 2.3.1.1 refers to an old and outdated Rev. O of the SSAR.
Westinghouse Response:
The ASI report has been modified to incorporate this comment.
(3)
Section 2.2.8, Main Feedwater Pumps, discusses the potential interaction between the a MFW Pumps the SGs. Of particular concern is continuing MFW flow after a reactor trip and overfilling or overcooling the SGs. The discussion in Section 3.4 (relating to Section 2.2.8) notes step 3 of ERG AES-0.1, " Check FW Status." Table 3-1, under Section 2.2.8 of the ASI report, indicates that the objective of this step is to instruct the closure of MFW control valves before establishing SFW flow to SGs (in step 3.c). However, step 3.a " response not obtained," in fact, bypasses closure of the MFW valves and goes directly to step 3.c. The objective of the step is not as clearly stated as Table 3-1 would imply. This ERG presentation could lead to higher potential for the noted error of commission.
Westinghouse Response:
The ERG background document will be revised to reference section 3 of the ASI report so that the insights captured in this section of the report are available to the COL applicant for thier use in developing the Emergency Operating Procedures.
(4)
Section 2.2.9, Startup Feedwater Pumps, addresses two interactions, SFW with PRHR and SFW during an SGTR event. The ASI report states that the PRHR can not be terminated until the SFW pumps are operational and water level in bolb SGs is 1
recovered. The ERGS (e.g., AES-1.1 step 6) conflict with the ASI report in that the i
termination criteria are SFW in operation and narrow range level in at least one SG l
recovered.
In addition, discussion in Section 2.2.9 on a SGTR event notes the problem with SFW causing an overfill and an SG PORV lift. The.ASI report states that SFW is i
6
9 automatically isolated on a high SG water level. This is not consistent with the l
ERGS. ERG AE-3, SGTR, does not mention SFW at all, and step 3.b tells the operator to stop feed flow. AFR-H.3, Response to High SG Level, step states " Isolate SFW to Affected SG." Neither one makes reference to the automatic isolation. Other places in the ERGS, where there are automatic isolations, have the operator " verify" the actions have occurred.
1 Westinghouse Response:
The ASI report has been revised to reflect the correct ERG termination criteria.
i The ERGS are consistent with the ASI report. Step 3 instructs the operators to stop feed flow to ruptured SGs In the ERGS, the term feed flow applies to feedwater supplied to the SGs from either the SFW or the MFW pumps / lines. However, the NRC is correct that the verification of the automatic isolation of FW isolation on high SG level is not included in ERGS AE-3 and AFR-H.3. Westinghouse will update the ERGS to include this verification as appropriate.
1 (5)
Section 2.2.13, Plant Control System (PLS), discusses how the PLS provides for the control and operation of the nonsafety-related systems. This section discusses the interactions that the PLS and pertinent nonsafety related systems have on the various i
accident scenarios analyzed in SSAR Chapter 15. These are summarized in Table 2.2 of the ASI report. Regarding section 15.4.6 of Table 2.2, it is noted that an adverse interaction is mitigated by automatic termination of boron dilution after a reactor trip.
However, AE-0, for a reactor response does not mention or verify this.
Westinghouse Response:
ERG AE-0 will be revised to incorporate this comment.
(6)
Section 2.2.14, Liquid Waste Processing System (WLS), states that the WLS collects and processes radioactive waste from the RCS, including from the ADS lines to the Reactor Coolant Drain Tank (RCDT). If ADS is actuated, RCS-V241 should automatically close to isolate the RCDT. A small adverse interaction is noted if the valve fails to close. The ERGS do not check this valve closed after an ADS actuation.
Westinghouse Response:
ERG AE-0 will be revised to address this comment. This verification has a low priority, and may be addressed with a generic note (as done in the standard ERGS) to inform the EOP i
writer where to add (or consider adding) steps to check appropriate operation of miscellaneous nonsafety related equipment as necessary (After Step 21 in AE-0).
I e
5 (7)
Section 2.2.16, IRWST Gutter, states the IRWST Gutter returns condensate from the containment walls to the containment sump normal operation, but realigns on PRHR actuation to return the condensate to the IRWST. This realignment is not checked in the ERGS.
i Westinghouse Response:
1 ERG AE-0 will be revised to address this comment.
t
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(8)
Section 2.3.1, Core Makeup Tanks (CMTs), deals with interactions of the CMTs with l
other pawive safety systems. It mentions operator action to isolate the accumulators before the accumulator tanks empty, in order to prevent injection of the nitrogen gas into the reactor. Neither AE-0 nor AE-01 contain a step to isolate the accumulators.
l l
Westinghouse Response:
t The ASI report has been modified to better explain that nitrogen injected into the RCS from the accumulators does not result in an adverse system interaction with the CMTs. Therefore, actions are not taken in the ERGS to attempt to isolate the accumulators prior to injecting nitrogen into the RCS.
(9)
Section 2.3.2, Accumulators, deals with interactions of the accumulators with other passive safety systems. ~ It also mentions operator action to isolate the accumulators before the accumulator tanks empty, in order to prevent injection of the nitrogen gas into the reactor. See comment (8) immediately above.
l Westinghouse Response:
l See the response to item 8 above.
(10)
Section 2.3 5, Passive RHR (PRHR), states that PRHR is very important for decay heat removal in an SGTR event. Yet ERG, AE-3 for SGTRs, in step 6.b isolates I
PRHR early in the ERG. These two facts seem contradictory.
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Westinghouse Response:
1 l
PRHR is not isolated "early" in the event. Prior to entering this ERG, the operator has l
already proceeded through AE-0, which includes verifying reactor trip and verification of the plant status, including proper operation of the safety systems and the nonsafety-related system. At that point, the operator transitions to the diagnosis phase of AE-0, (Steps 22-27) and transitions to the SGTR ERG. Not until Step 6 of this ERG (after which they identify o
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9 and isolate the faulted SG and verify the status of the RCS), does the operator verify the presence of a heat sink (SG level and SFW operable). At that time, it is appropriate to cool down the RCS with the steam dumps.
l In addition, the ERG background document will be revised to reference section 3 of the ASI report so that the insights captured in this section of the report are available to the COL applicant for thier use in developing the Emergency Operating Procedures.
Westinghouse will revise AE-3, Step 6 to clarify that the operator should verify level in the intact SG (as opposed to either SG) to preclude feeding a faulted SG.
(11)
Section 3.4, relating to Section 2.3.5.5, state that spurious actuation of PRHR affects the RCS by increasing reactor power. They state that the operators would follow ERG AE-0 and eventually in step 26 transfer to ERG AES-1.1, which in step 6 terminates PRHR. This appears to be an excessive delay in terminating the spurious actuation. Further, it is not clear that the spurious actuation would necessarily cause a safety injection signal which would be necessary to place one in AE-0 at step 26.
Westinghouse Response:
Termination of spurious actuations of safety-related systems may be performed by the operators without the use of procedures, prior to the generation of a plant trip or safeguards actuation signal. However, upon receipt of these signals, the operators are directed to follow the emergency operating procedures. The emergency procedures provide the explicit termination criteria for safety-related systems, once a reactor trip or safeguards actuation signal has been generated. This is consistent with the approach in current plants, with regards to spurious operation of safety-related systems.
(12)
Section 2.3.6.4, Automatic Depressurization System (ADS), discusses spurious ADS actuation. The staff's review of the ERGS has noted that the ADS is only addressed within the context of " verification of actuation / manual actuation backup." Unlike other " passive" safety systems (e.g., CMT) where the background documents make reference to spurious operation, the ERG background document information on ADS did not provide any guidance on isolation of ADS valves or actions subsequent to spurious operation of the ADS.
Westinghouse Response:
Operator actions to terminate inadvertent ADS are not included within the context of the ERGS. The resulting pressure transient due to inadvertent system level actuation of ADS would result in an immediate reactor trip and safeguard actuation (the entry point for the ERGS). Therefore, procedural actions would not be effective in preventing the consequences of an inadvertent ADS. Instead, the ERGS instruct the operator to take appropriate steps to ensure plant safety, and to mimmize the consequences of inadvertent ADS. The ERGS do 1
not address ADS termination criteria since this is dealt with via the long term action management plan.
In addition, since the entry point for the ERGS is a reactor trip or safeguards actuation, recovery of inadvertent operation of the CMTs and/or PRHR is not within the context of the ERGS. Termination criteria for these safety-related features are provided to recover the plant and transition to a normal condition.
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2.5.4 Data Display and Processing System Design Description The data display and processing system (DDS) provides nonsafety related alarms and displays, analysis of plant data, plant data logging and historical storage and retrieval, and operational support for plant i
personnel.
- 1. De DDS has distributed computer processors and video display units to support the data processing and display functions.
l 2.
The DDS provides for the minimum inventory of displays, visual alerts, and controls, as identified in Table 2.5.4-1,4 the R9ft M
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3.
The DDS provides information pertinent to the status of the protection and safety monitoring system.
4.
The DDS provides nonsafety.related displays of parameters originating in other systems.
Inspections Tests, Analyses,and Acceptance Criteria Table 2.5.4 2 specifies the inspections, tests, analyses, and associated acceptance criteria for the DDS.
1 2.5.4 1 j
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Table:1:4-a6 Minimum inventory of Displays and Controls 4erthe RfHt-A 5 W 1
I Description Control Display Alert'"
Neutron Flux N/A Yes Yes Neutron llux Doubling N/A No Yes Startup Rate N/A Yes Yes Reactor Coolant System (RCS) Pressure N/A Yes Yes Wide Range Hot Leg Temperature N/A Yes No Wide-Range Cold Leg Temperature N/A Yes Yes RCS Cooldown Rate Compared to the Limit Based on RCS N/A Yes Yes Pressure Wide Range Cold Leg Temperature Compared to the Limit N/A Yes Yes Based on RCS Pressure Change of RCS Temperature by more than 5' h in the last N/A No Yes 10 minutes Containment Water Level N/A Yes Yes Containment Pressure N/A Yes Yes Pressunzer Water Level N/A Yes Yes Pressunzer Water Level Trend N/A Yes No Pressunzer Reference Leg Temperature N/A Yes No Reactor Vessel Hot Leg Water Level N/A Yes Yes Pressunzer Pressure N/A Yes No Core Exit Temperature N/A Yes Yes RCS Subcooling N/A Yes Yes RCS Cold Overpressurs Limit N/A Yes Yes i
In-Containment Refueling Water Storage Tank (IRWST)
N/A Yes Yes Water Level f,
Passive Residual Heat Removal (PRHR) Flow N/A Yes Yes PRHR Outlet Temperature N/A Yes Yes 00hL I i OL L 2.5.4 3 W Westinghouse ())g? 1
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Table :.b.24(cont) d Minimum Inventory of Displays and Controls bhc RSR-MLd Description Control Display Alert'"
Passive Containment Cooling System (PCS) Storage Tank N/A Yes No Water Level PCS Cooling Flow N/A Yes No IRWST to Normal Residual Heat Removal System (RNS)
N/A Yes Yes Suction Valve Stan:s Demotely Operated Containment Isolation Valve Status
- N/A Yes No Containment Area HibbnRange Radiation Level N/A Yes Yes Containment Pressure (Extended Range)
N/A Yes No Containment Hydrogen Concentration N/A Yes No CMT Level N/A Yes No ManualReactorTrip k/srJr/A$h5 hrrf,,re trh)
Yes N/A N/A Manual Safeguards Actuation Yes N/A N/A Manual Core Makeup Tank Actuation Yes N/A N/A Manual Automatic Depressurization Systern (ADS) Stages Yes N/A N/A Ae3uokton
- 1. 2, and 3 l
....x Manual ADS Stage 4 we.;..a Ac.hserGon Yes N/A N/A Manual PRHR Actuation Yes N/A N/A Manual Containment Cooling Actuation Yes N/A N/A ManualIRWSTInjection Actuation Yes N/A N/A ManualContainment Recirculation Actuation Yes N/A N/A Manual Containment isolation Yes N/A N/A Manual Main Steam Line Isolation Yes N/A N/A Manual Feedwater Isolation Yes N/A N/A Manual Containment Hydrogen Igniter (Nonsafety Related)
Yes N/A N/A t (n.edeh-ReWek dbppp he % RSW))
n,,
(1) These parameters are used to gentrate visual alertbat identify challenges to the critical safety fu'nctions.
(2) These instruments are not required after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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Revision: 3 Effective: 4/18/97 Table 2.5.4-2 Inspections. Tests. Analyses, and Acceptance Criteria for the DDS Design Commitment inspections. Tests, Analyses Acceptance Criteria
- 1. The DDS has microcomputers Inspection of the as built system The as-built DDS has 4
I and video display units to support will be performed.
microcomputers and video display the data processing and display umts to support the data i
functions.
processing and display functions.
- 2. The DDS provides for the i) Inspection will be performed for i) The selected plant parameters minimum inventory of displays, retrievability of the selected plant can be retrieved in the RSR.
visual alerts. and controls, as paramete eMA the D
ii) Appropnate output signals are identified in Table 2.5.4 lpE ii) An operationaltest of the as-generated after the controls arthe
.MA
- 'd gnA,)
built system will be performed
-R$fure actuated.
using the controlsjrrtheitSit: %5W h
'ot
- 3. The DDS provides information Tests of the as built system will be The as built system pievides pemnent to the status of the performed.
displays of the bypassed and protection and safety monitoring operable status of the protection system.
and safety monitonng system.
- 4. The DDS provides nonsafety.
See Certified Design Material of See Certified Design Matenal of related disp!ays of parameters other systems.
other systems.
onginating in other systems.
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O' 5 cf Osn%d PROTECTION AND SAFETY MONITORING SYSTEM 7
Revision: 23 Effective: +0+SH964/18/97 b) n; P'IS pre. d;; for i; ;ran;f;r of;ca se: ;;peb!Hrj from i; ?.iCR ;;;h; r; ram; i doun comm ' RSR).
- n; P'!
- p. add;; f,,c i; m
- a mum in ca;;.> of disp:;3; and ;ca.rsl;. ;; :d;m;S;d la Tab l; l' M. in 1; RSR. n; ;ca:rol3 :n 6; RSR do na n;;d ;e M S:ad pcs.; ca.
a) D:;p:;5; aii; ap;al::a;;d a;=u;;ii; ru= r :e p br;;kr;;;a &...r.;.;d in i; ? CR.
8.
The PMS, in conjunction with the operator workstations, provides the following functions:
A a) The PMS provides for the minimum inventory of displays, visual alerts, and fixed position i
controls, as identified in Table 2.5.2-5, in the main cono on room (MCR).
T b) The PMS provides for the transfer of control capability t' rom the MCR to the remote shutdown
'g **at (RSRW u) u>cektblien c) Displa'ys of the open/ closed status of the reactor trip breakers can be retrieved in the MCR.
j l
- 69. a) The PMS automatically removes blocks of reactor trip and engineered safety features actuation when the plant approaches conditions for which the associated fhtion is designed to provide protection. These blocks are identific'd in Table 2.5.2 6.
b) TM I'.iS r..;;;..;;".y prod.;;; e r;;;;er ;r:p or ;agia;;r;d ;;fsj f;;;;;; !a;;;;;ca.per, r.
=;;mp;;c by pu cr.or; ir.. rec chr.c.a;;; of a f ac.c.. in.x; ree-c= ;,f fe,.s ;.::;.:.en lcg ;.
The PMS two-out-of-four initiation logic reverts to a twoout-of three coincidence logic if one of the four channels is bypassed. If a second channelis bypassed, the PMS two-out-of four initiation logic reverts to a one-out-of two coincidence logic. He PMS automatically produces a reactor trip or engineered safety feature initiation upon an attempt to bypass more than two channels of a function that uses two-out-of-four initiation logic N/ Qaared c1aane/t car-e olarme.A m de. McP.
3 c) The PMS provides the interlock tunctions identified in Table 2.5.2-7.
910.
Setpoints are determined using a methodology which accounts for loop inaccuracies, response testing, and maintenance or replacement ofinstrumentation.
- 10. n ; P'iS h i.r; r.r.d u,";a = ; ar;..r;G;d r.r.d vr.1 ir..;d ira f. a grow ;r. 1 y.
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y d;H..;d h a d r; :.:d ;; h.r;.
- 11. De PMS hardware and software is developed using a planned design process which provides for specific design documentation and reviews during the following life cycle stages:
5 a) Design requiremengphase hQ finition phase fordu> ace. WJ elopment phase se M * d)g est phase Gaten W Westinghouse
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D f JYfYlbf PROTECTION AND SAFETY MONI RING S STEM 8
Revision: 23 Effective * ' ^ '^ ~ /18/97 Table 2 5.2-5 Minimum Inventory of Displays and Fixed Position Controls, irl Ne MCR Description Control Display Alert'"
Neutron Flux N/A Yes Yes Neutron Flux Doubling N/A No Yes Startup Rate N/A Yes Yes Reactor Coolant System (RCS) Pressure N/A Yes Yes Wide Range Hot Leg Temperature N/A Yes No Wide Range Cold Leg Temperature N/A Yes Yes RCS Cooldown Rate Compared to the Limit Based on N/A Yes Yes RCS Pressure Wide Range Cold Leg Temperature Compared to the N/A Yes Yes Limit Based on RCS Pressure Change of RCS Temperature by More than 5'F in the last N/A No Yes 10 minutes Containment Water Level N/A Yes Yes Contamment Pressure N/A Yes Yes Pressurizer Water Level N/A Yes Yes Pressurizer Water Level Trend N/A Yes No Pressuruer Reference Leg Temperamre N/A Yes No Reactor Vessel-Hot Leg Water Level N/A Yes Yes Pressunzer Pressure N/A Yes No Core Exit Temperature N/A Yes Yes RCS Subcooling N/A Yes Yes RCS Cold Overpressure Limit N/A Yes Yes In-Containment Refueling Water Storage Tank (IRWST)
N/A Yes Yes l
Water Level Passive Residual Heat Removal (PRHR) Flow N/A Yes Yes l
PRHR Outlet Temperature N/A Yes Yes Passive Containment Cooling System (PCS) Stortge Tank N/A Yes No Water Level 2.5.2 9 W W65tingh00$8 mwsowtAAcsvww20502.viomas7 ame
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M PROTECTION AND SAFETY MONIT RING SY TEM I
I~I Revision: 23 Effective: 4ehW964/18/97 Table 2.5.2-5 (cont)
Minimum Inventory of Displays and Fixed Position Contronsa m Ne. MCA Description Control Display Alert"'
PCS Cooling Flow N/A Yes No IRWST to Normal Residual Heat Removal System (RNS)
N/A Yes Yes Suction Vaive Status Remotely Operated Containment isolation Valve Status'4 N/A Yes No Containment Area High Range Radiation Level N/A Yes Yes i
Containment Preaure (Extended Range)
N/A Yes No Containment Hydrogen Concentration N/A Yes No CMT Level N/A Yes No Manual ReactorTrip [4I:54 [f1/lihfes briiEt fr-Yes N/A N/A s
y Manual Safeguards Actuation Yes N/A N/A ManualCore Makeup Tank Actuation Yes N/A N/A Manual MCR Emergency Habitability System Actuation Yes N/A N/A Manual Automatic Depressurization System (ADS) Stages Yes N/A N/A 1,2, and 3 lemshes A e hv e4 6 /e a Manual ADS Stage 4inmenen Aeduo//ow Yes N/A N/A Manual PRHR Actuation Yes N/A N/A Manual Containment Cooling Actuation Yes N/A N/A ManualIRWST injection Actuanon Yes N/A N/A ManualContainment Recirculation Actuation Yes N/A N/A ManualContamment Isolstice Yes N/A N/A Manual Main Sesame Line isolation Yes N/A N/A Manual FeedweserIsolation Yes N/A N/A Manual Contamment Hydrogen Igniter (Noasafety.
Yes N/A N/A Related)
(Stjeh (t,M da)lE $ for t410 IOC y
Notes:
(
(l) These parameters are used so genersee visual eiens that idennfY chauenges no the endcal sasuty nacacas.
(2) These inseumens are not requesd aner 24 hows 2.5.2 10 WOS!Ingh0084 mwsowTAAcsvam20502.#Maoes7
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PROTECTION AND SAFETY MONITORING S TEM i
Revision: 23 i
Effective: +9eSH964/18/97 t
l Table 2.5.2-8 (cont)
Inspections. Tests. Analyses,and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria 4
l 38.a) na P' S r....d.; fu 6 i) Inspection will be performed for i) ne selected plant parameters
'c.;-...; M.....;., ;f d;.;:.,. xd retrievability of the selected plant can be retrieved in the MCR.
]
',.d p;
...;c...c.
.b,s.d....d parameters in the MCR.
i S T ib 21: 5. ' -h MCR.ne i
PMS provides for the minimum ii) Inspection willbe performed to ii) ne selected parameters are
{
inventory of displays, visual alerts, verify that the selected parameters used to generate visual alerts that l
and fixed position controls, as are used to generate visualalerts identify challenges to critical safety identified in Table 2.5.2 5. in the that identify challenges to critical functions.
I MCR.
safety functions.
j~
iii) An operational test of the as-iii) Appropriate output signals are built system will be performed generated after the MCR fixed i]
using the MCR fixed position position controls are actuated.
i controls.
18.b) The PMS provides for the An operational test of the as-built Actuation of the transfer switches d: b = ini.6:MC" "
R'5LA) transfer of con ol capability from system will be performed to re the activation of operator j
the MCR to the R6R-demonstrate the transfer of control 1
RgtO capability from the MCR to the control capability from thva';;.
i 96R-Riu) and the deactivation of operator control capability from the MCR.
l 78c) Oa."".00...;ds i i 0 ' :;; _ x ": h..'.. ad fr
- ) Oa
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j open/ closed status of the reactor
_.2,, t _ _ cni i i R6R-ereestussed-Displays of the trip breakers can be reineved in the
.i.".", - will be perfonned open/ closed staans of the reactor MCR.
for retnevaknisty of displays of the trip breakers can be reineved in the i
open/ closed staans of the teactor trip MCR.
breakers in the MCR.
i 69 a) ne PMS automatically An operational test of the as-built The PMS blocks are automatically I
removes blocks of reactor trip and PMS will be performed using real or removed when the test signal l.
engineered safety features actuation simulated test signals.
reaches the specified limit.
t when the plant approaches t
i conditions for which the associated I
function is designed to provide l
protection. Rese blocks are identified in Table 2.5.2 6.
2.5.2 15 W8Stingh0088 m wescowTAAcsvev3wo20$o2.#it440es7
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[
LJETSO/RM 468 EC EAST DATE. T!?E TO/FROM MODE MIN /SEC PGS CMDN STATUS 1
07 04/26 09:23 301 504 2222 G3-S 3 '29" 009 OK l
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l EMERGENCY RESPONSE FACILITIES
=
l Revision: 23 Effective: 4GhW964/18197 l
3.1 Emergency Response Facilities l
Design Description The technical support center (TSC) is a facility from which management and technical support is l
provided to main control room (MCR) personnel during emergency conditions. The operations support center (OSC) provides an assembly area where operations support personnel report in an emergency.
- 1. The TSC has floor space of at least 75 square feet per person for a minimum of 25 persons.
l l
- 2. The TSC has voice communication equipment for communication with the MCR, emergency operations facility, OSC, and the U.S. Nuclear Regulatory Commission (NRC).
- 3. The displays listed in Table 2.5.2-5, minimum inventory table, in s 0,5l~ I-Protection-ank-a v1 l
Safer, Mwudv..o- "pa.T. ("MS); can be retrieved in the TSC.
5 &eM (.Df 5)
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/
4.
The OSC has oic communication [quipment for co ication with the MCR and TSC.
Inspections, Tests, Analyses,and Acceptance Criteria Table 3.1-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the emergency response facilities.
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EMERGENCY RESPONSE FACILITIES Revision: 23 Effective: 496% 964/18/97 l
l l
Table 3.1-1 Inspections, Tests, Analyses,and Acceptance Criteris.
Design Commitment Inspections, Test Analyses Acceptance Criteria l
- 1. The TSC has floor space of An inspection will be The TSC has at least 1875 at least 75 square feet per perfonned on the floor space in square feet of floor space.
l person for a minimum of the TSC.
l 25 persons.
- 2. The TSC has voice An inspection and test will be Communications equipment is communication equipment for performed of the TSC voice installed, and voice communication with the MCR, communication equipment.
transmission and reception are emergency operations facility, accomplished.
2 OSC, and the NRC.
/
- 3. The displays listed in An inspection will be The disp s listed in Table'M minimum performed for retrievability of Table -
, minimum y l
inventory table, *
,Y the displays in th; TSC.
inventory table, in, l 45 t'-L subsection 2.
can be subsection 2.Se,
2,are NOe retrieved in the TSC.
retrieved in the TS.
- 4. The OSC has voice An inspection and test will be Communications equipment is communication equipment or performed of the OSC voice mstalled, and voice I
communication with the MCR communication equipment.
transmission and reception are and TSC.
\\
accomplished.
l 105 l
I 3.12 W Westinghouse oupeoovrAAesv vuosoi.wpeo4o997
l C:rtlfled Dxign Materlil
)
l HUMAN FACTORS ENGINEERING g -- --
Revision: 23 Effective: 40/34/06348474/18/97 l
l 3.2 Human Factors Engineering Design Description De AP600 human-system interface (HSI) will be developed and implemented based upon a human factors engineering (HFE) program.' Figure 3.2-1 illutrates the HFE program elements. The HST scope includes the design of the operation and control centers system (OCS) and each of the HSI resources. The OCS includes the main control room (MCR), the remote shutdown workstation (RSW),
the technical support center (TSC) and the operational support center. The HSI resources include the wall panel information system, alarm system, plant information system, computerized procedure l
system, soft and dedicated controls, and the qualified data processing system (safety related displays),
A minimum inventory of controls, displays and visual alerts are specified as part of the HSI.
l The main control room (MCR) provides a facility and resources for the safe control and operation of the plant. The MCR includes a minimum inventory of displays, visual alerts and fixed-position controls. Refer to item number 8.a. and Table 2.5.2-5 of subsection 2.5.2 for this minimum inventory.
De RSW provides the HSI resources to establish and maintain safe shutdown conditions for the plant l
from a location outside of the MCR. De RSW includes a minimum inventory of displays, controls and visual alerts. Refer to item number 2. and Table 2.5.4-1 of subsection 2.5.4 for this minimum inventory. As stated in item 8.b. of subsection 2.5.2, the PMS provides for the transfer of control capability from the MCR to the RSW.
1.
The integration of human reliability with human factors engineering design is performed m accordance with the implementation plan. Critical human actions (if any) and risk important j
tasks are identified and used as an input to the task analysis activities, j
2.
Task analysis is performed in accordance with the task analysis implementation plan... Task analysis identifies the irformation and control requirements for the operators to execute the tasks allocated to them.
3.
The HSI design is performed in accordance with the HSI design implementation plan. De HSI design includes the functional design of the operation and control centers and the HSI resources, the specification of design guidelines, the HSI resource design specifications, and the man in-the loop concept testing.
4.
n: CR : d th: :>:!!d!: MS! pe :!! :=tica cf 'iCR *:t by CR cpem'em te ep:mte S: p!::t : d =intin p!::t =fety.An HFE program verification and validation implementation plan is developed in y.cordance with the programmatic level description of the AP600 human factors verification 4nd validation plan. The implementation plan establishes methods for conducting evale.itions of the HSI design.
5.
The HFE verification and validation program is performed in accordance with the HFE verification and validation implementation plan and includes the following activities:
3.2 1
[ W85tiflgh00S8 O
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- TX CONFIRMATION REPORT **
AS OF APR 26 '97 09:36 PAGE.01 (JETSO/RM 468 EC EAST l
l DATE TIME TO/FROM MODE MIN /SEC PGS CMD# STATUS l
09 04/26 09:33 301 504 2222 G3--S 03'34" 004 OK i
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Certified Drign M;teri:1 e
HUMAN FACTORS ENGINEERING 5
Revision: 33 i
Effective: 444 406348474/18,57 a.
HSI Task support verification b.
HFE design verification c.
Integrated system validation d.
Issue resolution verification e.
Plant HFE/HSI (as designed at the time of plant startup) verification 6.
The MCR includes two reac:or operator workstations, one senior reactor operator workstation, safety-related displays, and safety-related controls.
7.
The MCR provides a suitable workspace environment for use by MCR operators.
8.
The h=en cy::er 5::rface (HSI) resources available to the MCR operators include the alarm system, plant information system, computerized procedure system, safety related displays, wall panel information system, and controls (soft and dedicated).
9.
The RSW includes one reactor operator workstation from which licensed operators perform remote shutdown operations.
10.
The remote shutdown room (RSR) provides a suitable workspace er~ironment, separate from the MCR, for use by the RSW operators.
11.
The HS1 resources available from the RSW include the alarm system displays, the plant information system, the computerized procedure system, and controls.
12.
The RSW and the available HSI permit execution of tasks by licensed operators to establish and maintain safe shutdown.
Inspection, Test, Analyses, and Acceptance Criteria Table 3.21 specifies the inspections, tests, analy'ses, and associated acceptance criteria for the HFE program, the MCR and the RSW.
l l
t 3.22 3 Westiligh0US8 mMp600VTAACSvev3Vt0302.wpf:1b.042597
C: stifled Dxign Material HUMAN FACTORS ENGINEERING is Revision: 33 W
Effective: 10,5?/^""/07'/18/97 f
Table 3.21 Inspections, Tests, Analyses, and Acceptance Criteria Design Comrnitruent inspections, Test, Analyses Acceptance Criteria
- 1. The integration of human An evaluation of the A report exists and concludes that reliability analysis with human implementation for the integration critical human actions (if any) factors engineering design is of human reliability analysis with and risk important tasks were performed in accordance with the human factors engineering design identified and examined by task implementation plan, will be performed.
analysis, and used as input to the HSI design, and procedure development.
- 2. Task analysis is performed in An evaluation of the A report exists and concludes that accordance with the task analysis impicmentation of the task function based task analyses were implementatiort plan.
analysis will be performed conducted in conformance with the task analysis implementation plan and include the following functions:
Control reactivity; control RCS boron concentration; control fuel and clad temperature; control RCS coolant temperature, pressure, and inventory; provide RCS flow; control main steam pressure; control SG inventory; control containment pressure and temperature; provide control of main turbine.
'J 3.23 W6SIh@t0US$
mfap600VTAACSvev39t0302.wpf:1b442597 l
Certlfled Declgn M; tert:1 HUMAN FACTORS ENGINEERING Revision: 33 l
Effective; 40 44/06348/074/18/97 Table 3.21 (cont)
Inspections, Tests, Analyses, and Acceptance Criteda Design Commitment Inspections, Test, Analyses Acceptance Criteda A report exists and concludes that operational sequence analyses (OSAs) were conducted in conformance with the task analysis implementation plan.
OSAs performed include the following:
- plant bestup and startup from post refueling to 100% power;
- reactor trip, turbine trip, and safety injection;
- natural circulation cooldown (startup feedwater with SG);
- loss of reactor or secondary coolant;
- post LOCA cooldown and depressurintion;
- loss of RCS inventory during shutdown;
- loss of RNS during shutdown;
)
- manual ADS actuation;
- manual reactor trip via PMS, via DAS; i
mode 1 l
l t
J v
3.24
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e Certlfled Dnign Material HUMAN FACTORS ENGINEERING i
Revision: 33 L
Effective: 4044963A38474/18/97 Table 3.21 (cont)
Inspections, Tests, Analyses, and Acceptance Criteria Design Corninitment Inspections, Test, Analyses Acceptance Criteria
- 3. The HS! design is performed An evaluation of the A report exists and concludes that in accordance with the HSI impicmentation of the HSI design the HSI design was conducted in design implementation plan.
will be performed.
conformance with the implementation plan and includes the following documents:
- Operation and Control Centers System Specification Document
- Functional requirements and design basis documents for the alarm system, plant information system, computerized procedure system, wall panel information system, soft controls, and the qualified data processing system.
- Design guideline documents for the Isim system, plant information system displays, computerized procedure system, and soft control displays.
- Design speci0 cations for the alarm system, plant information system displays, computerized procedure system, qualifted data procusiry system displays, wall panel information system displays, and controls (soft and dedicated).
- Man-in-the-loop concept test reports.
k[
f
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3.25
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Revision: 33 i
Effective. < n.,m. <.,inan,m.a.,in.,.,n18/97 I
1.
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Table 3.21 (cont)
Inspections, Tests, Analyses, and Acceptance Criteria l
l Design Commitment Inspections, Test, Analyses Acceptance Criteria 4."._'.'.'......"....."._.".."_..
T.._~..._"_....'.,"-^.'_
".. ' - ". -. "..... '., " ~.... - " _ '.. _
i HS! p:-d:== :!c: cf CP.
fel!=.t"n;; p!:-: r.'c! :!^= =d demens:=:: $.' 'h: CP ed: by Cn ep:=:c= te
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veriGcation and validation
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implementation plan is developed ep =:! ;; ch:: :t?~:= =d te ' ^^"; pe":-
in accordance with the
=p e== c' 6:.^.P'^^ d=:;;,
programmatic level description of
"" ' p:-fc :i
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the AP600 buman factors p!: ' te c!d de dete-verification and validation plan.
- 5) Mer=! p!- 5::'"; 2nd r.u-*";te !^^"; pc" n
!!) Brin;; $: p!=' te =fe n- $:cere! Sed de:det" '^!!:st a i
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- 6! = eip speeGed :=id=eA report exists and concludes that the HFB verification and validation iv) ^~:"--"-
implementation plan was
- =:!! ' =:k !ct c' :ce'-'
developed in accordar:ce with the acadens programmatic level description of
!=;;: b=d !cx cf :W*
the AP600 buman factors acedent verification and validation plan
~=="=b=d and includes (be foUowing
'= dte t: = b = d activities:
n!== ;;:=:::e- : !=
rup4wo. An inspection of the
- HSI Task support verification HFE verification and HFE design verification validation implementation
- Integrated system validation plan will be performed.
- Issue resolution verification
- Plant HFE/HSI veriGcation l
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HUMAN FACTORS ENGINEERING Revision: 33 Effective: 44/34/963/38/074/18/97 S. The HFE verification and A report exists and concludes l
validation program is performed that:
in accordance with the HFE veriGention and validation
- a. An evaluation of the
- a. Task support veris ation was implementation plan and includes implementation of the HSI task conducted in conformance with the following activities:
support verification will be the implementation plan and performed.
includes verification that the l
- a. HSI Task support verification information and controls provided
- b. HFE design verification by the HSI matebes the display l
- c. Integrated system validation and control requirements
- d. Issue resolution verification generated by the function based l
- c. Plant HFE/HSI(as designed at task analyses and the operational l
the time of plant startup) sequence analyses, verification
- b. HFE design verification was
- b. An evaluation of the conducted in conformance with implementation of the HFE design the implernentation plan and verification will be performed.
includes verification that the HS1 l
design is consistent with the l
AP600 specific design guidelines j
developed for each HSI resource.
j
- c. (i) An evaluation of the
- c. (i) "Ite test scenarios listed in j
implementation of the integrated the implementabon plan for l
system validation will be integrated system validation were performed.
executed in confonnance with the plan and noted human deficiencies were addressed.
- c. (ii) Tests and analyses of the
- c. (ii) The test and analysis following plant evolutions and results demonstrate that the MCR j
transients, using a facility that operators can perfonn the physically represents the MCR following:
configuration and dynamically i
represents the MCR HSI and the o Heat up and start up the i
operating characteristics and plant to 100% power responses of tbc AP600 design, will be performed:
o Shut down and cool down the plant to cold shutdown l
o Normal plant bestup and j
startup to 100% power o Bring the plant to safe l
shutdown following the o Normal plant shutdown and specified transients cooldown to cold shutdown o Bring the plant to a safe, o Transients: reactor trip and stable state following the turbine trip specified accidents 3.2/f l
T Westinghouse f,;
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J Certified Dicign M2t:ri;l j
HUMAN FACTORS ENGINEERING
- g=q l
Revision: 33 e
l Effective: 40/34063/34/074/18/97 l'
e 1
l I
Table 3.21 (c int)
Inspections, Tests, Analyses, at i Acceptance Criteria Design Commitment Inspections, Test, Analyw:
Acceptance Criteria
~
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l
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/9 / Hl f Accidents:
g Uli9
- small break loss-of-coolant
/
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- targe-break loss-of-coolant accident I!)/g i
steam line break i
b8 feedwater line break steam generator tube rupture
- d. An evaluation of the
- d. HFE design issue resolution l
implementation of the HFE design verification was conducted in issue resolution veriGeation will conformance with the be performed.
implementation plan and ingtudes veriGestion that human factors issues documented in the design issues tracking system have been addressed in the final design.
l l
- e. An evaluation of the
- e. The plant HFE/HSI, as l
implementation of the plana designed at the time of phnt HFE/HSI(as designed at the time startup, is consistent with the of plant startup) verification will HFE/HS! verified in Sa. through be performed.
Sd.
d) i U
3.2-)df h 3 W85tingh0US8 m :'ap600VTAA CSV ev3Vt0302.wpf:1 b 442597
o Certlfled Declgn Material HUMAN FACTORS ENGINEERING
(
Revision: 23
+
Effective: 1C ??"f"/07'/18/97
- 6. The MCR includes two An inspection of the MCR The MCR includes two reactor reactor operator workstations, one workstations and control panels operator workstations, one senior senior reactor operator will be performed.
reactor operator workstation, workstation, safety-related safety related displays, and displays, and safety-relatrd safety related controls, controls.
- 7. The MCR provides a suitable i) See Certified Design Material, i) See Certified Design Material, workspace environment for use subsection 2.7.1, Nuclear Island subsection 2.7.1, Nuclear Island by the MCR operators.
Non-radioactive Ventilation Non-radioactive Ventilation System.
System.
ii) See Certined Design Material, ii) See Certified Design subsecr%n 2.2.5, MCR Emergency Material, subsection 2.2.5, MCR Habitability System.
Emergency Habitability System.
iii) See Certified Design Material, iii) See Certified Design Material, subsection 2.63, Class IE de and subsection 2.6.3, Class IE de and UPS System.
UPS system.
- 8. The HS! resources available to An inspection of the HS!
The as-built HSI includes an the MCR operators include the resources available in the MCR alarm system, plant information alarm system, plant information for the MCR operators will be system, computerized procedure system, computerized procedure performed.
system, safety-related displays, system, safety.related displays, wall panel information system, wall panel information system, and controls (soft and dedicated).
and controls (soft and dedicated).
- 9. The RSW includes one reactor An inspection of the RSW will be The RSW includes one reactor operator workstation from which performed, operator workstation, licensed operators perform remote shutdown operations.
- 10. The remote shutdown toom See Certified Design Material, See Certified Design Material,
(RSR) provides a suitable subsection 2.7.1, Nuclear Island subsection 2.7.1, Nuclear Isisad workspace environment, separate Non radioactive Ventilation Non radioactive Ventilation from the MCR, for use by the System.
System.
RSW operators,
- 11. The HSI resources available An inspection of the HSI The as-built HS1 at the RSW at the RSWinclude the alarm resources available at the RSW includes alarm system displays,,
system displays, the plant will be peformed.
plant information system, information system, tbc computerized procedure system, computerized procedure system, and controls.
and controls, w--
---_ m
4 Certifled D:cign M;terial s
HUMAN FACTORS ENGINEERING Revision: 33 F#
Effective: 4044/063/as/974/18/97
- 12. The RSW and the available Test and analysis, using a A report exists and concludes that l
HSI permit execution of tasks by workstation that physically the test and analysis results licensed operators to establish and represents the RSW and demonstrate that licensed maintain safe shutdown.
dynamically represents the RSW operators can achieve and HSI and the operating maintain safe shutdown characteristics and responses of conditions from tbc RSW.
the AP600, will be performed.
l l
u 3 2 Y jb W westinghouse mwmumesv.em.e*2m
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00 TX CONFIRMATION REPORT cc AS OF APR 26 '97 09 45 PAGE.01 LETSO/RM 468 EC EAST DHTC T!ME TO/FROM MODE M!WSEC PGS CMD# STATUS
.10 04/26 09:37 301 504 2222 G3-S 07'29" 009 OK ememen I
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Westinghouse Energy Systems ecx 355 Electric Corporation Pmsburgh Pennsyhania 15230 0355 i
DCP/NRC0829 -
Docket No.: SW 52 003 '
Fax Note r.
767; g
April 25,1997 -
Document Contre IlS W
7 g
i U. S. Nuclear Re co
,u Ah.r j
Washington, DC e%
Medeg-V7(~n g l
ATTENTION: T.R. QUAI
SUBJECT:
AP600 DESIGN CERTIFICATION; FORMm..
)N OF RESOLUTION OF ITEMS ASSOCIATED WITH SECTION 9.4.
l
References:
1.
SECY-97 051, " Schedule for the Staffs Review of the AP600 Design l
Cenification Application," dated February 26,1997, forwarded by NRC letter
. " Westinghouse's Support of the Nuclear Regulatory Commission Review of the AF600 Design Certification Review," dated March 6,1997.
l 2.
Letter, T. R. Quay to N. J. Liparuto, Subject, " Staff Update to Open items (Ols) and Request for Reinstatement of Deleted information Regarding Section 9.4 of the Westinghouse AP600 Standard Safety Analysis Repon (SSAR)," dated October 17,1996.
Dear Mr. Quay:
This letter is to formally consolidate responses and resolutions ofitems associated with SSAR Section L
9.4 and to confirm completion of submittal of final documentation related to SSAR Section 9.4 for our application for AP600 Design Certification. Reference I includes a milestone " Applicant Submits Final SSAR Revision & Documentation" by May 1997. Westinghouse interprets this to require NRC acknowledgement of receipt of final documentation supporting our application for AP600 Design Certification. To support this milestone, NRC and Westinghouse maintain a' detailed activity plan-p.
which provides schedule goals for most SSAR/FSER sections and related activities, such as, the PRA, code validation, and ITAACs. In this detail activity plan, Westinghouse application input and NRC internal FSER input for Section 9.4 of the SSAR has a schedule goal of May 15,1997. NRC and j
Westinghouse also maintains a joint open item tracking system to informally monitor the status and history of open items (DSER, RAI, meeting, and other) associated with our application.
L '
This letter with its attachments provides answers and resolutions to the questions and comments'in the Reference 2 letter. Attachment I to this letter provides the chronology for each item discussed. Some of the material responding to the Reference 2 letter has been sent previously and is referenced in
- p.. Attachment 2 provides marked up pages incorporating proposed changes to Section 9.4 m
0
a NSD-NRC-97-5087 DCP/NRC0829 2
April 25, 2997 of the SSAR resulting from the items identified on Attachment I and incorporates recommendations of the Reference 2 letter as well as other proposed changes. Attachment 3 is the marked up pages for Section 9.4 that are affected by our design change decision to include the Radiation Chemistry Laboratory subsystem within the Auxiliary / Annex Building Subsystem. The changes involved in both Attachments 2 and 3 will be included in Revision 12 of the SSAR.
j Based upon responses p,ovided for Section 9.4 of the SSAR and a review of the related open item I
entries in our informal tracking system, Westinghouse confirms its completion of the submittal of information to support this portion of our application.
If you have any comments or questions on this letter please contact J. W. Winters (412-374 5290) or l-D. A. Lindgren (412-374-4856).
B- /#'
Brian A. McIntyre, Manager Advanced Plant Safety and Licensing i
jml : Chronology for Responses and Marked Up Pages for Section 9.4 of the SSAR : Marked Up Pages for Section 9.4 of the SSAR : Markup for Incorporation of Rad Chem Lab Subsystem into the Auxiliary / Annex Building Ventilation Subsystem l'
cc:
D. Jackson. NRC N. J. Liparuto, Westinghouse (w/o Attachments)
]
T. T. Martin, NRC (w/o Attachments) i i
?
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1 n t:4 4
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h Sheet 2 of 5 CHRONOLOGY OF RESPONSES AND MAR;GD UP PAGES FOR SECTION 9.4 OF THE SSAR ITEM ANSWERS MATERIAL IS IN NUMBER IN OITS SSAR, REV.
II, MARK UP PAGES OF SSAR, NRC LETTER NUMBER TYPE DATE PAG E-REVISION 11, PROMISED 5.e.
FAX 1/16/97 p.4-53, -70 f
NSD-NRC-5.f.
293 97-4932 1/7/97 f
5.g.
FAX 1/13/97 9.4-45 l 4-9, -14, -15 12/27/96 j
5.h.
FAX 12/31/96 9.
6.a.(1)
MARK UP OF PAGE 9.4-5 6.a.(2) 9.4-10,-11 6.a.(3)
MARK UP OF PAGES 9.4-10,-11,-12 t
6.a.(4) 9.4-1 I e
NO CHANGE REQUIRED. SYSTEM ISOLATES MAIN CONTROL ROOM ON APPROPRIATE RADIATION 6.a.(5)
SIGNALS 6.a.(6)
MARK UP OF PAGE 9.4-13 6.a(7)
MARK UP OF PAGE 9.4-14 NO CHANGES REQUIRED - THIS IS UNNECESSARY DETAIL FOR THE 6.a.(8)
SSAR NO CHANGES REQUIRED - THIS IS UNNECESSARY DETAIL FOR THE 6.a.(9)
SSAR f
[
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Sheet 1 of 5 ATTACHMENT 1 TO NSD-NRC-97-5087 CHRONOLOGY OF RESPONSES AND MARKED UP PAGES FOR SECTION 9.4 OF THE SSAR 1
ITEM ANSWERS MATERIAL IS IN NUMBER IN OITS SSAR, REV. II, MARK UP PAGES OF SSAR, NRC LETTER NUMBER TYPE DATE PAGE-REVISION II, PROMISED 1.a.
TABLE 9.4 NEW SHEETS 6 & 7 l
1.b 9.4.2.2.1 TEXT PAGES ONLY 1.c.
TABLE 9.4 3 NEW SHEETS 2 & 3 9.4-2 PLUS FIG. 9.4.2-1 NEW SHEETS 3 1.d.
&4 1.e.-I NO FIGURE REQUIRED 9.4.8 IS ON PAGE 9.4-50 AND 9.4.11 IS 1 e.-2 ON PAGE 9.4-67 9.4-17,-19,-28,-30,
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FAX 2/20/97
-51,-57, -68,-81 l
2.b.
264 FAX 2/21/97 5.4-15 3.a.
FAX 1/26/97 9.4-43 3.b.
FAX 11/26/96 9.4-56 HARD 9.4-56 AND i
4.
COPY I1/18/96 TIIROUGHOUT 9.4 l
5.a.
MARK UPS OF PAGES 9.4-7, -8, -45, -46 5.b.
FAX 1/15/97 p.4-53, -58, -70 I9.4-9,-21, -32, -46, i
5.c.
FAX I/15/97
-53,-64,-70 5.d.
FAX 1/13/97 9.4-20 l
3
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Sheet 3 of 5 CHRONOLOGY OF RESPONSES AND MARKED UP PAGES FOR SECTION 9.4 OF THE SSAR ITEM ANSWERS MATERIAL IS IN NUMBER IN OITS SSAR, REV. 11, MARK UP PAGES OF SSAR, NRC LETTER NUMBER TYPE DATE PAGE-REVISION 11, PROMISED 6.b.(1)
MARK UP OF PAGE 9.4-17 NON-IE BATTERY ROOMS DON'T 6.b,(2)-1 HAVE ELECTRIC REHEAT COILS 6.b.(2)-2 9.4-18 NON-1E BATTERY ROOMS DON'T 6.b.(3)
HAVE ELECTRIC REHEAT COILS 6.b.(4)
MARK UP OF PAGE 9.4-25 6.c.(1) 9.4-28 6.c.(2)
MARK UP OF PAGE 9.4-30 6.c.(3) 9.4-33 6.d.(1)
MARK UP OF PAGES 9.4-43,-48 6.d.(2) l MARK UP OF PAGE 9.4-48 NO CHANGES REQUIRED - TIIIS IS UNNECESSARY DETAIL FOR THE 6.d.(3)
SSAR 6.e.(1)
NO FIGURE REQUIRED 6.e.(2)
MARK UP OF PAGE 9.4-54 6.e.(?>
MARK UP OF PAGE 9.4-54 6.f.(1) i NO FIGURE REQUIRED 7.a.(1) 2897 FAX 12/9/96 fARIOUS IN 9.4 l
7.a.(2) 261 FAX 12/9/96 f.4-7
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r Sheet.4 of 5 CHRONOLOGY OF RESPONSES AND MARKED UP PAGES FOR SECTION 9.4 OF THE SSAR l
s ITEM ANSWERS MATERIAL IS IN NUMBER IN OITS SSAR, REV.
11, MARK UP PAGES OF SSAR, NRC LETTER NUMBER TYPE DATE PAGE-REVISION 11, PROMISED 7.a.(3)
FAX 1/17/97 l9.4-8, -7I B 9.4-2, -10, -11, AND -
7.a.(4) 2890 FAX 12/17/96 6.4-1 TECII. SPEC. 3.7.6
& SSAR TABLE 3.9-I 7.a.(5) 263 FAX 12/6/96 17
[
7.a.(6)
FAX 12/9/96 9.4-11 i
7.a.(7) 264 FAX 2/21/97 9.4-15 7.b.(1) 270 9.4-17,-19 7.b.(2) 269 FAX 4/8/97 RESOLVED 7.b.(3)-1 271 FAX 4/8/97 MARK UP OF PAGE 9.4-17 j
7.b.(4) 272 RESOLVED 2/21/97 i
7.b.(5) 266 FAX 4/15/97 MARK UP OF PAGE 9.4-19 7.b.(6) 273 FAX 12/9/97 9.4-19 h
7.b.(7) 267 FAX 4/7/97 9.4-19 7.b.(8) 274 FAX 4/7/97 9.4-19 7.c.(1) 284 FAX 12/10/96 9.4-33 7.c.(2) 281 TABLE 3.2-3 7.c.(3) 282 FABLE 3.2-3 t
i
SHEER 5 of 5 CHRONOLOGY OF RESPONSES AND MARKED UP PAGES FOR SECTION 9.4 OF THE SSAR ITEM ANSWERS MATERIAL IS IN NUMBER IN OITS SSAR, REV. 11, MARK UP PAGES OF SSAR, NRC LETTER NUMBER TYPE DATE PAGE-REVISION 11, PROMISED
,19.4 - 3 1 &
7'.c.(4) 275, 276 FAX 12/11/96 l TABLE 3.2-3 7.c.(5) 279
'9.4-30 7.c.(6) 283 TABLE 3.2-3 7.d.(1) 285 FAX 4/11/97 TABLE 3.2-3 7.d.(2) 287 FAX 12/26/96 19.4 - 3 9 7.e.(1) 291 FAX 4/16/97 9.4-43 7.e.(2) 289 FAX 4/11/97 TABLE 3.2-3 4/10/97 7.f.(1) 292, 293 FAX 4/16/97 TABLE 3.2-3 t
7.f.(2) 294, 295, 2 %
'9.4-51
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9.4-57 &
l 7.g.(1) 298 FAX 4/16/97 TABLE 3.2-3 i
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1 I/9/97 7.h.(2) 300 FAX 4/16/97 MARK UP OF PAGE 9.4-62 j
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1/28/97 i
7.i.(1) 304 FAX 4/16/97 FIG. 9.4.4.7-1 7.i.(2) 302. 305 TABLE 3.2-3 i
i
- TX CONFIRMATION REPORT **
AS OF APR 29 '97 13:29 PAGE.01 APG00 DESIGN CERT DATE TIME T0/FROM MODE MIN /SEC Cy STATUS 01 4/29 13:27 1823 NRC G3--S 02'05 03 OK
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AUTHOR DATE CH K'D. BY DATE VERIFIED SY DATE
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j Westinghouse Revised Response to ASI Discussion Item #4 I
Discussion Item #4 1
L The discussion on fan coolers (pp. 2-17'- 2-18) appears to be focused on an extreme situation without regard fo'r other potential adverse conditions. For example; recent concerns were
)
raised regarding operating PWRs via Westinghouse's Nuclear Safety Advisory Letter on containment fan coolers. Are there any high heat load conditions (such as during DBA or Severe Accidents) for the AP600 in which the cooling water system supplying non-safety-
- related fan coolers (Chilled Water) might be subject to water hammer or other potentials for containment bypass? Could operation of the fan coolers with chilled water isolated by a l-
- containment isolation signal result in overpressurization of the chilled water line or flash-ing/ water hammers if the heated chilled water lines from the fan coolers were suddenly unisolated?
l l
The Chapter 9 SSAR description of the fan coolers state that they have two speed motors.
The high speed is used for normal conditions and the low speed is used during high containment air density conditions - such as those that might be present during DBA or severe accidents. How is fan speed controlled during accident conditions. Since this is a non-safety related function, how is operation of the fans in fast speed prevented in a high steam l
environment. Does the fan control circuitry automatically shift to low speed in accident i
conditions? Are there interlocks to prevent operators from manually shifting to high speed when conditions may be inappropriate in containment. Is there any potential for the fans to catastrophically fail if operated at high speed in a dense steam environment thereby' creating a l
l.
possible adverse system interaction which could damage the chill water cooling coils l
(Containment bypass scenario)? The emergency response guidelines (ERGS) for reactor trip or safety injection AE-0, step 22, does not specify at what speed the fans should be operating under safety injection conditions.
i L
. Westinghouse Response:
Westinghouse has performed an evaluation to assess the design and operation of the AP600 L
fan coolers and chilled water system following an accident, with regard to the concerns raised -
about fan coolers for operating plants in NRC Generic 12tter 96-06. The following is a summary of that evaluation.
The AP600 has containment fan coolers which remove heat from the containment during normal operations. These fan coolers are cooled b'y chilled water supplied by the central chilled water system (VWS). The AP600 fan coolers and VWS are not relied on for safety-
- related containment cooling. The only safety function of the VWS is to isolate the chilled water supply and return lines to the containment recirculation system (VCS) fan cooler coil units, following any event resulting in a containment isolation signal to provide containment 4
j integrity. During cold weather when the plant is shut down, the high capacity subsystem is also designed to permit use of the chilled water piping inside containment for containment 3
~-
. - -. - - - - _.~ -. - -
heating.. Manual realignment of the system allows hot water to be supplied from the hot water heating system (VYS) to the VCS fan coils units. Therefore, the chilled water piping inside containment is insulated.
I De postulated accident scenario for the evaluation is a double-ended guillotine break of one of the reactor coolant system cold legs. This event results in a rapid increase in containment pressure (and temperature) to near the containment design pressure, and in the longer term containment pressure and temperature can remain elevated. As a result of the increase in containment pressure and/or the loss of reactor coolant the chilled water supply and return piping from the fan coolers is isolated almost immediatly. In the AP600 design the operating fan cooler fans continue to operate at low speed, circulating the heated containment atmosphere across the cooler coils. This would cause the stagnant chilled water in the coolers
. to heat-up to and be maintained at the containment air / mixture temperature. If the VWS piping is subsequently unisolated after the water in the coolers has been heated, the heated water in the coolers may flash. De subsequent supply of cold water from the VWS to the coolers, could collapse the steam creating a water hammer. Also of concern is that flashing could occur in the coolers when they are in operation at elevated containment temperature conditions, causing two-phase flow and loss of cooling flow capability. Unlike other PWR plants, the VCS and VWS in the AP600 are not required to operate post-accident and they are not restarted automatically. The purpose of the evaluation given in this report is to determine if and when, in view of the GL96-06 water hammer concerns, the VWS system can be used i
post accident.
The Westinghouse evaluation has concluded that sufficient overpressure exists such that the AP600 chilled water system can be unisolated and operated with no flashing of chilled water, provided that the containment temperature is below 230 F. This evaluation provides the following precautions and limitations to prevent flashing and potential water hammer in the chilled' water piping:
1
- Following an event which results in heat up of the containment air / steam above 230*F,
'the isolated cooling water supply and return containment isolation valves should not be opened to restore chilled water flow to the operating fan coolers, until the containment atmosphere temperature has been reduced to s 228*F.
- Following an event which results in heat up of the containment air / steam above 230*F,
. cooling water flow should not be initiated to fan cooler coils unless the fans for these coolers has been running for a sufficiently long time to ensure the water in the coils is at equilibrium temperature with the containment atmosphere temperature, and until the containment atmosphere temperature has been reduced to s 228'F.
- The chilled water flow to operating fan coolers should be stopped and isolated using the containment isolation valves, whenever the containment atmosphere temperature exceeds 230*F.
- Following an event which results in heat up of the containment air / steam above 230*F, l
chilled water flow to operating fan coolers should be initiated by first opening the L
l N
chilled water return line isolation valves before the supply line isolation valves.
The evaluation concludes that if these precautions and limitations are adhered to, the chilled water piping inside containment will not be subject to water hammer, that could lead to a containment bypass scenario. Overpressure protection of the chilled water system is provided by two thermal relief valves connected to the inlet and outlet headers. These relief valves prevent the chilled water piping design pr:ssure to be exceeded following containment isolation, and subsequent heatup of the containment. with the chilled water system pipimg water solid. Another design feature of the chilled water piping and containment fan cooler coils is the ability of these systems to withstand a perfect vacuum. Following the containment heatup and cooldown postulated in the evaluation, the resulting minimum pressure in the system piping and components can approach a perfect vacuum, and the system piping and components shall be designed to accomodate this condition.
The containment fan coolers are equipped with two-speed fans. The high speed is used for normal conditions, and the low speed is used primarily to perform the containment integrated leak rate testing, that requires the containment to be pressurized to design pressure.
Normally, the fans are manually re-aligned to perform this test. However, following a transient or accider.t that results in the containment pressure and temperature to be elevated, the fan me, tors receive a signal to automatically switch to low-spped operation, based on a pre-set containment pressure signal. This is accomplished via the nonsafery-related plant control system, in addition, if the fans were to operate in a high containment pressure /
temperature condition, the fan motors are provided with a thermal overload switch that would automatically trip the motor to prevent damage. Operation of the fans at high speed with high containment pressure could damage the fan motors, but would not cause damage to the chilled water system.
This response has been incorporated into section 2.2.12 of the ASI report. The AP600 cmergency response guidelines background documents will be updated to incorporate the precautions and limitations associated with restarting the chilled water system following ar.
accident.
D
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3.4.10 i
i i.
3.4 REACTORCOOLANTSYSTEM(RCS)
(-
3.4.10 RCS Leakage Detection Instrumentation LCO 3.4.10 The following RCS leakage detection instrumentation shall be OPERABLE:
a.
One containment sump level channel; b.
One containment atmosphere radioactivity monitor (gaseous N13/F18).
APPLICABILITY:
MODES 1, 2, 3, and 4.
i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME t
A.
Required containment
NOTE--------------
5 rs J.V,is i
sump channel LC0 3.0.4 is not applicable.
4.2.
l
[
A.1 Perform SR 3.4.8.1 (RCS Once per inventory balance).
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND j
l A.2 Restore containment sump 3o A$r3 l
channel to OPERABLE 15" br; I
status.
I i
l (continued) i l-l l
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h AP600 3.4-15 08/96 Amendment 0 i
miwn mie --
o fp
~
ncs Leanege vetection instrumentation 3.4.10 l
l '.-.
(I ACTIONS (continued) l l
CONDITION REQUIRED ACTION COMPLETION TIME l
l B.
. Required containment
N0fE-------------
l atmosphere LC0 3.0.4 is not applicable.
g 3,y,fy radioactivity monitor inoperable.
8' E B.I.! Analyze grab samples of Once per containment atmosphere.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OR B.1.2 Perform SR 3.4.8.1.
Once per i
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l
AND Jo Drff3 B.2 Restore containment 159 heers atmosphere l
radioactivity monitor l
to OPERABLE status.
fhours C.
Required Actica and C.1 Be in MODE 3.
i associated Completion Time not met.
AND
/ 2-C.?
Be in MODE 4.
N hours v
AWO c.3 /u,7,47a Acnou r, s z. p,<5 RS57ME we!!44/t&
I LE4ch5F DE7Ec-7/oA'
.msre,wvemw ro oPr#dte tr47us.
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l, m.a ecou3e ve cec c un insu u.nent.6 un 3.4.10 l
a-b SURVEILLANCE REQUIREMENTS l
l SURVEILLANCE FREQUENCY l
/2-SM SR 3.4.10.1 Perform a CHANNEL CHECK of required 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> gg3,4,/S,1 l
containment atmosphere radioactivity monitor.
C6T SR 3.4.10.2 Perform an C"""1 OI: PATI ^"AL TE&T of 92 days required containment atmosphere radioactivity monitor.
SR 3.4.10.3 Perform a CHANNEL CALIBRATION of 24 months required containment sump monitor.
SR 3.4.10.4 Perform a CHANNEL CALIBRATION of 24 months required containment atmosphere
(
radioactivity monitor.
l h AP600 3.4-17 08/96 Amendment 0 AP01)sensusest10030410 AHOSJD6 9
e FAX to DINO SCALETTI April 28,1997 i
CC:
Sharon or Dino, please make copies for:
Diane Jackson Ted Quay Don Lindgren 1
Bob Vijuk Brian McIntyre OPEN ITEM #1102 (DSER 9.3.6-3) 1
\\
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
l Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 35 calendar days away (25 business days). The relevant documentation related to Open Item ##1102 (DSER 9.3.6-2) deals with the i
non-safety-related chemical and volume control system. (See attached OITS material.), therefore, the criteria specified in Section 9.3.4 of the Standard Review Plan do not apply to this system.
It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed."
o#
Jim Winters 412-374-5290 i
/ /t
- AP600 Open Item Tracking System Database: Executiva S - ~ry
- Date: 4/28/97
.l Selecties:
[ item no] between 1102 And 1102 Sorted by item #
l Item DSER Section Titic/ Description Resp (W)
NRC Brwy QQ _,
. _ _ ype- -
__,_g__1 Swn Engh Status Swm T
o
,n_-__
,, _, @_ Np; f _, _., Dase i102 NRR/EMCB 93 6-3
. DSER-OI Winters Closed
' Action W i
{ Westinghouse should address enteria iden't'ifEd (Section 93.4 oldSRP'cMktiN CES.[
~ ' ~
~
]
[
(ClosedITh'e AP600 chemic~al'and volwneNt'rol" syms not a safety relased syssern. ThercfA,~thicAecria listed in Ection 91[of the SRP is'
~
~
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inot applicable s' ce they address only safety relased systems.
m N
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i Page: 1 Total Records: I m
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FAX to DINO SCALETTI April 28,1997 CC:
Sharon or Dino, please make copies for:
Ted Quay j
Don Lindgren Bob Vijuk Brian McIntyre OPEN ITEM #1101 (DSER 9.3.6-2) l To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 35 calendar days away (25 business days). The relevant documentation related to Open Item ##1101 (DSER 9.3.6-2) deals with the non-safety-related chemical and volume control system. (See attached OITS material.) Since there is no regulatory treatment of non-safety related systems, it is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status I
l of this item. We recommend " Action N" or " Closed."
Jim Winters 412-374-5290 l
l l
i
//t
s.
AP600 Opco Itzm Tracking System Datbase: Executive Summary
'Date: 4/28/97 -
- Selecties:
litem nol between i101 And 1101 Sorted by item #
Item DSER Section Titic/Descriptior, Resp (W)-
NRC No.'
. Branch Question
-Type Detail Status Engineer Status Status letter No. / -
Dane., _
1101-NRR/EMCB 936-2 DSER4M Winters Closed Actkm W
' Westinghouse must satisfactorily address th'eTssue of the regulkM of M-rela'acd syst5fdthe chemical and solunic ontrol~
~
~
^
c
. ys_g as wel[as pelieryism-safety-relased systems. __
s
[CloE{The AP600 chemical andIoieme mnerol syseem has no 151 SI'M functions.
_ _ _. [_,
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i FAX to DINO SCALETTI April 29,1997 CC:
Sharon or Dino, please make copies for:
Diane Jackson Ted Quay Don Hutchings Don Lindgren Bob Vijuk Brian McIntyre OPEN ITEM #3088 (RAI 410.279)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 32 calendar days away (24 business days). The relevant docmnentation related to Open Item #3088 (RAI 410.279) is in a Westinghouse letter to NRC, NSD-NRC-96-4765, which provided the Westinghouse response to this RAI. Pertinent pages of this letter are attached (these were sent to you in July of 1996 -
- over 9 months ago). It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed."
Jim Winters 412-374-5290 W >
NRC REQUEST FOR ADDITIONAL INFORMATION i
l Question 410.279 Why were the curbs around the sumps used to keep out debris removed from the design in Revision 4 to SSAR Section 9.3.57 l
Response
SSAR subsection 9.3.5.2.2, Revision 7, states that sumps are covered to keep out debris. Covers are removable or manholes are provided for access. In addition, sumps may have other features to aid in minimizing accumulation of debris. For example, depending on location and type of sump, some have curbs or special debris traps or both.
SSAR Revision: NONE l
l
[
I l
l d
410.279-1 i
I
i NSD NRC-96-4765 DCP/NRC0549 July 1,1996
. Correspondence with respect to the application for withholding should reference AW-96-984 and should be addressed to Brian A. McIntyre, Manager of Advanced Plant Safety and Licensing, Westinghouse Electric Corporation, P.O. Box 355, Pittsburgh, Pennsylvania, 15230-0355.
Please contact Brian A. McIntyre on (412) 374-4334 if you have any questions concerning this transmittal.
J YvS Brian A. McIntyre, Manager Advanced Plant Safety and Licensing
/nja Enclosures Attachments cc:
T. Kenyon, NRC (w/o enclosures / attachments)
W. Huffman, NRC (IEl, IE2)
R. C. Jones, NRC (w/o enclosures / attachments)
G. D. McPherson, NRC (w/o enclosures / attachments)
.I F. Eltawila, NRC (w/o enclosures / attachments)
R. Landry, NRC (IEl)
P. Boehnert, ACRS (4EI)
N. J. Liparuto, Westinghouse (w/o enclosures / attachments) t f
)-
n>
4 O
e
x AP600 Open Iteen Tracking Systems Datbasei Exce:tiva Seamanary '-
Dates 4/29/97 Selecteen:
{ item no] between 3088 And 3088 Sorted by Type h
DSER Section
. Title / Description Resp (W)
NRC No.
Branch Question Type Detail Status Engineer Status 1 Status letner No. I Dune 3088 NRR/SPLB 9.15 RAI4M Winsers
' Closed Action W - NTD-NRC-96-4765 i
[410179 Why were the curbs around the surnps used to keep out debris removed from the design in Revision 4 to SSAR Section 9.3.5?
]
- {CE.Reshpdide[via~iFestingiumseleuer NS5 RF964763[datedi 19%.
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Energy Systems ecoss g
Electric Corporation; eswo rennsyiono5230 c333 -
l.
6 NSD-NRC-96-4765 DCP/NRC0549 Docket No.: STN-52-003 July 1,1996 Document Control Desk U.S. Nuclear Regulatory Commission l
Washington, D.C.' 20555 ATTENTION:
T.R. QUAY
SUBJECT:
WESTINGHOUSE RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION ON THE AP600
Dear Mr. Quay:
l.i Enclosed are three copies of the Westinghouse responses to NRC requests for additional information on AP600. ' Responses to RAls on the OSU Scaling Report, long term cooling analysis and the SSAR.
- are included in this transmittal. The specific RAls are listed on the attachment to this letter.
The NRC technical staff should review these responses as a part of their review of the AP600 design.-
i These responses close the 20 RAls.
The Westinghouse Electric Corporation copyright notice, proprietary information notice, application for withholding and affidavit are attached.
This submittal contains Westinghouse proprietary information consisting of trade secrets, commercial information or financial information which we consider privileged or confidential pursuant to 10CFR2.790. Therefore, it is requested that the Westinghouse proprietary ' formation attached in m
L Enclosure i be handled on a confidential basis and be withheld from public disclosure. The non-proprietary copy'of Enclosure 1 is provided as Enclosure 2.
This material is for your internal use only ard may be used for the purpose for which it is submitted.
L
!! should not be otherwise used, disclosed, duplicated, or disseminated, in whole or in part, to any
[
' other person or organization outside the Commission, the Office of Nuclear Reactor Regulation, the i
Office of Nuclear Regulatory Research and the necessary subcontractors that have signed a proprietary -
non-dischsure agreement with Westinghouse without the express wrinen approval of Westinghouse.
i py> -
1 2
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FAX to DINO SCALETTI April 28,1997 CC:
Sharon or Dino, please make copies for:
Diane Jackson Ted Quay Don Hutchings Bob Vijuk Brian McIntyre OPEN ITEM #1112 (DSER 9.4.11-1)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 35 calendar days away (25 business days). The relevant documentation related to Open Item #1112 (DSER 9.4.11-1 is in my fax to Diane Jackson sent January 28,1997 and in SSAR Figure 9 4.4.7-1. Peninent pages of these documents are attached and were re-sent to you with my fax of April 16,1997 regarding Open Item 304. (Other items dealing with this same subject are Items 302,303, and 305 which are either " Closed" or " Action N.") It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this item. As shown on the attached, we now recommend " Action N" or " Closed."
Jim Winters 412-374-5290 s
h6
~
+.
- AP600 Open Itesn Tracking Systesa Database: Executiva Sonnenary Date: 4/2857 Selectiea:
htem nol between ii12 And ii12 Sorted by Item # -
hem DSER Section Title / Description Resp.
(W)
NRC
. No.
Branch Question Type Detail Status Engineer _
Status Status W No. I Dese til2 NRR/SPLB 9.4.11-1 DSER-OI Winters Closed Action W J %has get hnn{
Mi of,hfspo_
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. {5 sed - All mquested infM has 5 providedith' NRC.
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Page: 1 Total Records: I
FAX to DIANE JACKSON January 28,1997 This is in response to item 7.i.(1) of your 10/17/96 letter, item 7.i.(1) of our 12/17/96 telecon and OITS item 304. HEPA filtration is not required on the VHS exhaust to the plant vent because the high radiation alarm in the exhaust line would lead the operators to terminate discharge on high radiation. This approach is consistent with the VRS and the two VAS exhaust paths to the plant vent shown on Figure 9.4.7-1 (Sheet 1 of 2). We recommend that the "NRC Status" for item 304 be changed to " Action td l
Thanks Jim Winters 412-374-5290 cc:
McIntyre Cummins Hutchings 3h
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- 9. Auxillay Systims
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Contmimnent Air Filtration System Piping and Instrumentation Diagram Revision: 7 3 W85tiflgh0tlS8 April 30,1996 l
k 9.4-14 1
FAX to DINO SCALETTI April 28,1997 CC:
Sharon or Dino, please make copies for:
Diane Jackson Ted Quay Don Hutchings Don Lindgren Bob Vijuk Brian McIntyre OPEN ITEM #3085 (RAI 410.276)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 33 calendar days away (25 business days). The relevant documentation related to Open Item #3085 (RAI 410.276) is in a Westinghouse letter to NRC, NSD-NRC-9f727, which provided the Westinghouse response to this RAI. Pertinent pages of this letter are attached (these were sent to you in May of 1996 -
over 1I months ago). It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed."
s\\'
Jim Winters 412-374-5290
i AP600 Ope:2 Item Tracking Syrtem Database:' Exec tiv Sur.amary Dat : 4/28/97
[
Selecties:
litem nol between 3085 And 3085 Sotted by Type item DSER Section Title / Description Resp (W)
NRC I
No Branch Question Type Detail Status Engineer Status Status grt.t_er No. I.__
. _D._ ate _.
3C*.5 NRR/SPLD 94 RAl-OI Winters Closed Action W NSD-NRC-96-4727 f 410.276 SSAR Section 9 4.1, Nuclear Island Nonradioactive Ventilation System Table 9 4.1-1 of the SSAR identifies assumed in-leakages thmugh the Main Control Room (MCR) access doors and the MCRRechnical Suppat Center (TSC) equipment ductwork (operating) and out leakages through the MCR structure and the through MCRRSC lleating, Ventilation and Air
- Conditioning (ilVAC) equipment and ductwork (operating). Westinghouse should state that during abnormal operation with high suborne radioactivity conditions, the MCRRSC llVAC subsystem can limit the doses to the control room operators to General Iksign Criteria (GDC) 19 I
dose limits given the assumed in-and out-leakages _ _ _ _ _ _
,I' Closed - Response pro'ided by
'.it 4R69U72f
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i Westinghouse Energy Systems eenass Electric Corporation Pit'SDufin Pennstvania 15230 0355 NSD-NRC-96-4727 j
DCP/NRC0517 i
Docket No.. STN-52-003 1
t May 20,1996 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION:
T.R. QUAY
SUBJECT:
WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION ON THE AP600
Dear Mr. Quay:
Enclosed are three copies of the Westinghouse responses to NRC requests for additional information on the AP600 topics. Responses to RAls 410.276,410.278,410.284,410.287,410.290,410.292, 410.293,410.294,440.354,440.355,471.24, and 952.98 are attached in this transmittal.
The NRC technical staff should review these responses as a part of their review of the AP600 design.
These responses close the RAls.
j l
l Please contact Brian A. McIntyre on (412) 374-4334 if you have any questions concerning this transmittal.
A l
Brian A. McIntyre, a ger Advanced Plant Safety and Licensing
/nja Enclosures j.
cc:
T. Kenyon, NRC (w/o enclosures)
D. Jackson, NRC W. Huffman, NRC C. Li, NRC l
L. Lois, NRC N. Liparuto, Westinghouse (wlo enclosures) l
,.pl
l S.
i l
l NRC REQUEST FOR ADDITIONAL INFORMATION
- =u Ouestion 410.276 Re: SSAR Section 9.4.1, Nuclear Island Nonradioactive Ventilation System Table 9.4.1-1 of the SSAR identifies assumed in-leakages through the Main Control Room (MCR) access doors and the MCR/ Technical Support Center (TSC) equipment ductwork (operating) and out leakages through the MCR structure and'the through MCR/TSC Heating. Ventilation and Air Conditioning (HVAC) equipment and ductwork (operating). Westinghouse should state that during abnormal operation with high airborne radioactivity conditior.s.
the MCRfTSC HVAC subsystem can limit the doses to the control room operators to General Design Criteria (GDC) 19 dose limits given the assumed in and out. eakages.
?
Response
The " Abnormal Plant Operation" portion of SSAR subsection 9.4.1.2.3.1 (Main Control Room / Technical Support Center HVAC Subsystem). Revision 7. states that the rates shown in Table 9.4.1 1 " maintain operator doses within allowable limits." The GDC 19 portion of SSAR subsection 3.1.2. Revision 7. explicitly states that shielding and HVAC design limit doses in the main control room to less than the required 5 rem whole-body, or its equivalent.
during the accident.
SSAR Revision: NONE i
i l
1 W Westingflouts
./
410.276-1 l
1 l
4
[
i e
FAX to DINO SCALETTI l
April 28,1997 CC:
Sharon or Dino, please make copies for:
Diane Jackson Ted Quay Don Lindgren Bob Vijuk Brian McIntyre OPEN ITEM #1458 (DSER 19.2.2.1-4)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 33 calendar days away (25 business days). The relevant documentation related to Open Item #1458 (DSER 19.2.2.1-4) is contained in Sub-Section 9.5.1 and Appendix 9A of the SSAR. This material was sent to you with Revision 8 of the SSAR in July of 1996 (about nine months ago). !t is requested NRC review this material and provide def'mitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed."
Jim Winters l
412-374-5290
\\
l 1
l i
1
()z-1
p AP600 Open Iteam Tracking System Database: Executive Somemary -
Date: 4/28M7 Selectme:
litem no] between 1458 And 1458 Sorted by Type
' Im DSER Section Title / Description Resp (W)
NRC No
'Ikanch Question 3
Detail Status Engineer Status Status T pe -
letter. No. / __... D_ ate.'..
1458 NRR/SCSB 19.2.2.1-4 D5ER-OI Winters Closed Action W -
[Westingiumse should addre(s the remlut'ionif fireMEM (([_ _
]~.
~
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IClosc5SSAR subsection 9.3~l 'and Apped9A5si5n 8, provide detail ofth5Isrc protecti$n'shtem and remtv'c nowii NRfconcenis.~Herc~
fwe no outstanding NRC requests for addnional infonnation on fire protection._
+
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Page: 1 Total Records: I u.-
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FAX to DINO SCALETTI April 28,1997 CC:
Sharon or Dino, please make copies for:
Diane Jackson Ted Quay l
Don Lindgren
'#~
Bob Vijuk Brian McIntyre OPEN ITEM #1763 (DSER 1.10-1)
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all Westinghouse submittals by May 30,1997. This is just 33 calendar days away (25 business days). The relevant documentation related to Open Item #1763 (DSER 1.10-1) is contained in Sub-Section 1.8 of the SSAR. More detailed action items for the COL are listed in applicable sections of the SSAR. It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed."
'(
Jim Winters 412-374-5290
/,[ 2.
AP600 Open Itesa Tracking Systems Database: Executive Seammisry Date: 4/28/97 Selection:
litem nel between 1763 And 171,3 Sorted by Type leese DSER Section Titic/ Description Resp (W)
NRC No Branch Question Type _...
Engineer Status Status Detail Status Lette.r_No. / _.. _D_at.e_. _
1763 NRR/PDST 1.10-1 DSER-Of Winters Closed Action W COL applicant and hcensees whoAfe5 tine ~ cst'ifed'APtds~ standard Asign in the futurim~ill be'rchiredAsatisfy the requiremess and ~
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,e in the design control document (DCD). The staff has identified certain requirernents and s.
- in the DSER. These COL action items related to programs, pis.
o, and issues that are outside the scope of the certified design review. Westinghouse is expected to identify an acceptable list of COL action items in the SSAR and DCD.
[
Closed - A lis't of'M~Iicense indion items was established in Chapter I as part ofiiectEll Mo' re'dc5iledempectatEfor each E r
'~
f are included in the appropnate sections of the SSAR.
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~. _. _ _ _ _ _ _. _ _
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- TX CMIRr% TION REPORT **
AS OF CPR 30 '97 08:50 PAGE,01 g
LJETSO/RM 468 EC EAST l
DATE TIME TO/FROM MODE MIN /SEC PGS CMDu STATUS l
12 04/30 08:42 301 504 2222 G3-S 03'26" 006 001 OK 13 04/30 08:46 516 344 4 G G3-5 03'38" 006 001 OK A
h W Westinghouse FAX COVER SHEET AECIPIENT INFORMATION
$ ENDER INFORMATION I
CATE:
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NAME:
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'Til AsdaAss h L f)'Haes p,,h -3')fpl j, g, u car 10N:
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PHONE:
Aff L-S PHONE:
CCMPANY:
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NA/L.
j LOCATION:
FAX:,
(4121 374 5099 l
l f
Cover
- Pages J
MEMOVE ALL STARES PENCIL WILL NOT TRAN5MIT. USE SLACK PEN PLEASE MAKE COPtES OF TWQ SIDED PAGE5 f
Comments:
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1.0 INTRODUCTION
I
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,Mrs am R
This document provides a programmatic level description of the AP600 Human Factors Verification j
and Validation (V&V) plan. It specifies at a high-level the activities to be performed as part of the l
AP600 V&V. Individual implementation plans that provide more detailed descdptions of the tests to
('
be performed, and acceptance criteria to be used, will be developed for each V&V activity specified in this report. Individual V&V implementation plans will be developed after design certification.
\\
1.1 AP600 V&V Activities and Objectives
'Ihe Human Factors Engineering Program Review Model (PRM) developed under the sponsorship of the U. S. NRC (NUREG-0711) specifies that an HFE V&V program should include five activities with the following objectives:
/
Verifies that the h.un nn ty:dern f
" - interfac,: :.. OMilS,
- 1. Task Support Venfication:
design provides all necessary ala displays, and controls to support plant personne! tasks,
- 2. HFE Design Verification:
Verifies that the M-MIS design conforms to human factors engineering (HFE) principles, guidelines, and standards
- 3. Integrated System Validation:
Validates that the M-MIS design can be effectively operated by personnel within all performance requirements
- 4. Issue Resolution Verification:
Verifies that the M MIS design resolves all identified HFE issues in the tracking system
- 5. Final Plant HFE Verification:
Verifies that the final as-built procuct conforms to the verified and validated design that resulted from the M-MIS design process The AP600 V&V will include all five of these activities. Figure 1-1 presents the AP600 V&.V activities and sequence in which these activities shall be performed. The sequence for completing these V&V activities will be as follows:
05[
1.
M-MIS-Task Support Verification 2.
HFE Design Verification l
3.
Integrated System Validation i
4.
Issue Resolution Verification 5.
PlantMMerification 4
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ts being operated to show that these tasks can be accomplished without interfering with operator tasks j
i necessary for monitoring and controlling the plant j
4.6 Criteria for Selection of Test Scenarios for Dynamic Evaluadons L
' A multi-dimensional set of criteria will be used to define a set of test scenarios to be included in the integrated system validation. Dimensions to be considered will include covering:
l A range of operational modes including normal plant evolutions (startup, full power, and shutdown) i Transients (reactor trip, turbine trip)
Design-basis and beyond design-basis accidents covered by the EOPs AP600-specific design features (the Automatic Depressurization System, the Diverse Actuation System)
L Scenarios that include human perfonnance actions identified to be risk important by l'
the PRA 1
Instrument failures M-MIS equipment and processing failures, including failure of the computerized l
procedure system, establishing the ability to use the back up 4
Reactor shutdown and cooldown from remote shutdown panel
=
Situations that produce cognitive challenges, including situations that complicate:
Situation assessment by providing degraded or conflicting plant state information l
Response (require balancing of multiple goals, require maaal takeover of l
I automanc systems) 1 Performance by increasing personnel communicadon/coordmation f
requirements or I
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Increase workload by introducing additional tasks or distractions 3
(Subsection 4.5 & 4.7) r The set of test scenarios specified will be sufficient to validate the EOPs as implemented in
- computerized procedures.
They will also include scenarios to validate key HRA modeling assumptions for event sequences that involve risk.imponant human actions. Examples of assumptions to be confhmed are that particular human actions that need to be performed are satisfactorily completed within the time-window specified
. in the PRA.
1 The set of test scenarios included in integrated system validation will be defined by a j
multi-disciplinary team that includes input from EOP developers, M-MIS designers, human factors specialists, and human reliability analysis /PRA analysts.
666 N Ser + 4.7 Realisde Validation Scenarios i
The implementation plan will specify how test scenarios will be realistic with respect to plant conditions that are likely to hold for the simations being represented (number of perscanel in the control room, commumcation requirements with personnel outside the control room, requirements for notification to outside organizations, noise level and temperature).
Selected scenarios will include environmental conditions, such as noise and distractions, which may affect human performance in an actual nuclev power plant.
For actions outsioe the connel room thu are within the scope of the integrated system vahdation, performance impacts of potentially harsh environments that requus additional time will be reahstically simulated (for example, time to dm protective clothing and access bot areas).
4.8 Performance Measures and Acceptance Criteria The implementation plan will specify performance measr$s used to establish that mission goals and operator performance reqirements are achieved. Performance measures will include:
System measures relevant to plant safety Personr.el pnmary task performance i
Personnel enors i
I Situation awareness l
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6.0 S'.% PLANT """ C:0[ VERIFICATION v
An implementation plan will be developed specifying a methodology for verifying that the t
7 :rennrn,- in th, M mis d.<ign ihme =nlead from the HFE design process and V&V activities.
F6//H1 (b levreA O TW h dt fyld $ W Q In the Westinghouse design process, mee'hanisms for insuring that systems conform to the final functional requirements and desip descriptions, are factory acceptance tests conducted on the actual system hardware at the factory, and the site acceptance test conducted after the hardware is installed at the plant site.
The implementation plan for th erification will specify the verifications that will be conducted as part of the factory acceptance test, and site acceptance test, ensuring that the dant conforms to the M-MIS desip that resulted from the IGE desip process and V&V tfFE-/#sIle in p/> M-g h4e d /kd tb}gy activities.
J The implementation plan will include procedures for identifying aspects of the M-MIS that were not addressed in the design process V&V, and procedures for evaluating them using appropriate V&V methods. Aspects of the M-MIS desip that fall in this category 'melude desip features that couid not be evaluated in a simulator, and desip modifications that occurred subsequent to the M-MIS desip V&V, such as hardware upgrades.
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T To:
Bill Huffman - NRC
Subject:
NRC/W NOTRUMP Meeting Date:
April 17.,1997 Pages:
Four, including this cover sheet.
COMMENTS:
l l
- Bill, I
i Attached is marked-up copy of your draft letter. Call me if you have any quesders. Thanks.
1 I
I l
l L. Hochreiter (FAX), B. Osterrieder, M. Young, cc:
B.MCINTYRE (NRC Informal Correspondence prom in. oesa of...
File), A.Gagnon, File 7.6
,,,,g,,,,,,,,,,,
l Manager, Advanced and WER Plant Safety Analysis Westinghouse PO Box 355 Pittsburgh, PA 15235 (412)374-4790 Fax: (412) 374-5744
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l APPLICANT: Westinghouse Electric Corporation PROJECT:
AP600
SUBJECT:
SUM ARY OF MEETING TO DISCUSS AP600 NOTRUMP SMALL BREAK LOSS COOLANT (LOCA) CODE VERIFICATION AND VALIDATION The subject meeting was held on March 13, 1997, in the Rockville, Maryland, offices of Westinghouse Electric Corporation between representatives of Westinghouse and, the Nuclear Regulatory Commission (NRC) staff. The purpose of the meeting was to discuss overall conclusions from the AP600 NOTRUMP computer code final verification and validation, additional documentation which will be included in the revised final verification and validation report, and information to be presented to the ACRS.
Highlights from the meeting include the following items:
.+
Westinghouse noted that based on a QA check ;5:t calc notes, some corrections would be made to the NOTRUMP final V1V.
The SPES 2-in::h cold leg balance line break (501007) simulation did not accurately reflect the test break location and will be revised in the final V&V report. Westinghouse presented calculations that showed no significant changes in the simulation results with the corrected break location.
The final V&V will be corrected.
Westinghouse stated that for the SPES test comparisons, the PRHR loop is removed after ADS-3 to enhance the running time. Westinghouse also stated that NOTRUMP tends to underpredict PRHR heat transfer but that this only. impacts small break simulations ( je.u h 4.%
y,14 J ;j[
NOTRUMP tends to predict delayed ADS actuations relative to the tests at
/'
both SPES and OSU.
The noding and models used for 05U, SPES, and AP600 calculations are consistent and differences are due to facilitygence;.
In general, for the OSU simulations, NOTRUMP underpredicts ADS flows and overpredicts break flows to the extent that total mass inventory tends to be consistently underpredicted.
NOTRUMP simulations of small breaks (less than 2 inches) are not predicted as well as larger breaks because of the underprediction of a
PRHR heat transfer and lack of a thermal stratification model for the CMT.
i I
i APR 17 #97 11:57 301 504 2300 PAGE.02
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! Westinghouse also discussed some documentation issues in the final V&V and how j
they would be. addressed. The changes Westinghouse intends to make are summarized below:
j i
the SPES mixture level plots will be revised to account for variable area fluid modeling.
i
[
The comparison of AP600 NOTRUMP calculation with volumetric flow-based momentum equations versus mass ficw-based momentum equations will be j
removed from the final V&V (section 3.5).
Westinghouse stated that the
= final version of the NOTRUMP code cannot be switched between volumetric and mass flow-based versions and that all calculations are now done with
' volumetric flow-based momentum equations.
J A a I v,4 % %..s
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Westinghouse developed a quench del for NOTRUMP for cases where core nodes uncover and then recover. Westinghouse has concluded that the quench model is not needed for AP600 NOTRUMP calculation and has not included a description in the final V&V report. However, Westinghouse i
noted that some of the G-2 benchmarking problems required the use of the quench model due to output spiking and run failures. Westinghouse still plans to include those G-2 plots where the quench model was used. The staff recommended that plots of the G-2 cases with and without the quench model be included in the final V1V and that Westinghouse clearly state that the quench model will not be used for design basis analysis of the AP600.
The staff had stated in.the SOSER that the NOTRUMP levelizing model was only used in some limited cases. Westinghouse noted that the final version of NOTRUMP will use the levelizir.9 model
- & all h.e. w i-t AwIm.
Preparation for an upcoming presentation to the ACRS on the NOTRUMP computer code was also discussed.
Westinghouse is preparing a summary background report on NOTRUMP which it will issue for the ACRS prior to the meeting.
The meeting was productive and progress towards resolution of the remaining issues was made. Westinghouse agreed to the following actions (which would be documented in the Open Item Tracking System) as a result of discussion during i
the meeting:
i Copies of the NOTRUMP related RAI responses and the background summary report being prepared for the ACRS will be included in the updated NOTRUMP final V&V report.
Responses to RAIs 440.215 and 440.216 will be incorporated into the text of the final V&V.
Similar to the documentation for the NOTRUMP final V&V, Westinghouse M f::
' i+ include a background summary and all appropriate RAI responses in the LOFTRAN CAD.
- APR17[9711:58 301 504 2300 PAGE.08.
"i_.3;.-5; -25;;
3' l is the list of meeting attendees.
l is a copy of the presentation handouts with material removed which Westinghouse claims is proprietary.
Westinghouse committed to submit via separate correspondence an application for withholding, affidavit, and non-proprietary copy of the proprietary presentation material.
A draft of this meeting sunmary was provided to Westinghouse to allow them the opportunity to ensure that the representation of comments and discussion was accurate.
William C. Huffman, Project Manage" Standardization Project Directorate Division of Reactor Program Management Office Of Nuclear Reactor Regulation Decket No.52-003 Attachments: As stated l
cc w/atts: See next page l
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APR 17 '97 11:58 301 504 2300 PAGE.04
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l FAX TO DINO SCALETTI April 28,1997 l
CC:
Sharon or Dino please make copies fer :
Ted Quay l
Don Lindgren Richard Orr Bob Vijuk l
Brian McIntyre l
[
OPEN ITEM #1975 (DSER 19.2.6.4-1)
- 1888 (DSER 3.8.2.4-1)
- 3271 (RAI 220.102) i l
To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Resisions &
Documentation" by 5/97, we believe that NRC must acknowledge receipt of all l
Westinghouse submittals by May 30,1997. This is just 35 calendar days away (25) business days). The relevant documentation related to the three open items covering the same subject matter, #1975 (DSER 19.2.6.4-1), #1888 (DSER 3.8.2.4-1),and #3271 (RAI I
220.102) is in the SSAR, Sub-Section 3.8.2.4.2.5. (pertinent pages are attached). This material was submitted to you in Revision 7 of the SSAR on May 6,1996 (more than 11 months ago).
Additional information on these open items was sent to you by letter NSD-NRC-97-4981 on February 11,1997 (pertinent pages attached). It is requested NRC review this material and provide definitive action for Westinghouse or provide direction to change the status of these three items. We recommend " Action N" or " Closed."
A Jim Winters 412-374-5290 l
i.
l i
- AP600 Open Itema Tracking Systema Database: Exec tivaSummary Date: 4/25/97 Selection-
[ item no! between 1975 And 1975 Sorted by Type hem DSER Section Title / Description Resp (W)
NRC
- No.
Branch.
Question Typc Detail Status Engineer Status Status LeuctNo.I Date 1975 NRR/ECGB 19.264-1 DSER-COL Winters Closed Action W l19.2.6.4-1 ' ' 'The electrical penetration assemblics to be used shall be at least as strong as the steel containment vessel.
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l' age: 1 Total Records: I
_ _ _ _ _ _ _ _ - _ _ _ _ _ _ = _ _.
AP600 Open Item Tracking System Dat; base: Executive Summary Date: 4/25/97 Selection:
[ item no] between 1888 And 1888 Sorted by Type item -
DSER Section
~
Resp -
.(W)
NRC Titic/ Description No.
Branch Qsestion Type _
-.. _ -. - ~ - -..
Engineer Status Detail Suus.
Status
'1 tNo
-._ - - _ _ - -. x_a. c. ---.. s- - -
Danc...-
1888 NRR/ECGB 382.4-1 DSER-COL Orr Closed Action W NSD-NRC-97-4981 tme A'sh y d Ee _,
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Disidssefin meetind UBI 8/3073I55. ExMCOL information to inUInde demonAat5on that EPA satisfies 5EEevel U pressuriand
~
temperature requiremer:t. Revise SSAR 3 8 6.1 to change *ultimaec capacitics* to
- ultimate pressure capacities
- Closed: additional clarification is requested under RAI 220.102 transmined by NRC letter dated April 4,1996.
{losed - Response provided by NSD-NRC-97-4981 of 2/11/97.
-b Page: 1 Total Records: I
AP600 Open itemi Tracking Systen Database: Execctive Su:nniary Datst 4/25/97 Selecties:
[ item no] between 3271 And 3271 Sorted by Type item -
DSER Section Titic/ Description Resp
' OW)
NRC No Branch Question Typc Detail Status Engineer Status Status
_ _ --._ett_er. _N_o_. /.-... Da_te__
L 3271 NRR/ECGB 3.82 RA14)t Cll42K)rr/Lutz Closed Action N NSD-NRC-97-4981
~
~
220.IO2 1n SSAR Section 3 821.E.5,'nMicaiand clectncal penetrEEare desigrEd [E~a pressure of 90 psig's design EnMurel280 F) fd ASME Service Level C limits. In SSAR Section 3.8.2.4.2.8, however, the ASME Service Level C limit is 92 psig at 280 F from the
(*!.I.
kad p,_ stgh{ ling. Mfy wpich pressure represents the ASME Service Level C lirnit at design temperature for the contanunent..
a Closed - Response provided by NSD-NRC-97-498I of 2/11/97.
-}
Action N - Based upon E-Mail 'AP600 OITS* from Scalcui on 2/2697._
.l Page: 1 Total Records: I
+ -., -
- 3. Design of Stnsctures, Components, Equipment, and Systems head is taken as the theoretical plastic buckling pressure of 174 psig predicted in the BOSOR-5 analyses.
The deterministic severe accident pressure capacity is taken as 60 percent of critical buckling.
This is consistent with the safety factor for Service Level C in ASME Code, Case N-284 and results in a containment head capacity of 104 psig.
3.8.2.4.2.3 Equipment Hatches l
SECY 93-087 permits evaluation of certain severe accident scenarios against ASME Service 1
Level C limits. The equipment hatch covers were evaluated for buckling against ASME I
paragraph NE-3222 and according to ASME Code, Case N-284. Use of ASME Code, Case N-284 for this application was confirmed to be approprit.te by ASME. The containment intemal pressure acts on the convex face of the dished head and the hatch covers are in compression under containment internal pressure loads. The critical buckling capacity is based on classical buckling capacities reduced by capacity reduction factors to account for the effects of imperfections and plasticity. These capacity reduction factors are based on test data and are generally lower-boend values for the tolerances specified in the ASME Code.
The critical buckling pressures are 195 psig for the 22-foot-diameter hatch and 160 psig for I
the 16-foot-diameter hatch at an ambient temperature of 100*F. For the Service Level C l
limits in accordance with paragraph NE 3222, a safety factor of 2.50 is specified, resulting I
in capabilities of 78 psig (22-foot-diameter) and 64 psig (16-foot-diameter). For the Service l
Level C limits in accordance with Code Case N284, a safety factor of 1.67 is specified, resulting in capabilities of 117 psig (22-foot-diameter) and 96 psig (16-foot-diameter).
Typical gaskets have been tested fc severe accident conditions as described in NUREG/CR-5096 (Reference 25). The gaskets for the AP600 will be similar to those tested with material such as Presray EPDM E 603. For such gaskets the onset of leakage occurred at a temperature of about 600 F.
3.8.2.4.2.4 Personnel Airlocks The capacity of the personnel airlocks was determined by comparing the airlock design to that tested and reported in NUREG/CR-5118 (Reference 3). Critical parameters are the same, so the results of the test apply directly. In the tests the inner door and end bulkhead of the airlock withstood a maximum pressure of 300 psig at 400*F. The capacity of the airlock is therefore at least 300 psig at ambient temperature. The maximum pressure corresponding to Service level C is conservatively estimated by reducing this capacity in the ratio of the minimum specified material yield to ultimate.
i 3.8.2.4.2.5 Mechanical and Electrical Penetrations Subsections 3.8.2.1.3 through 3.8.2.1.6 desc:ibe the containment penetrations. Penetration reinforcement is designed following the area replacement method of the ASME Code. The insert plates and sleeves permit development of the hoop tensile yield stresses predicted as the r
l l
I f
Revision: 11
[ Westirighotise j
3.8-13 February 28,1997 l
l
- 3. Design of Structures, Components, Equipment, and Systems limiting capacity in subsection 3.8.2.4.1.
Capacities of the equipment hatch covers are discussed in subsection 3.8.2.4.2.3 and of the personnel airlocks in subsection 3.8.2.4.2.4.
Mechanical penetrations welded directly to the containment vessel are generally piping systems with design pressures greater than that of the containment vessel. Thicknesses of the flued head or end plate are established based on piping support loads or stiffness requirements.
The capacities of these penetrations are greater than the capacity of the containment vessel cylinder.
i Mechanical penetrations for the large-diameter high-energy lines, such as the main steam and feedwater piping, include expansion bellows. The piping and flued head have large pressure capability. The response of expansion bellows to severe pressure and deformations is described in NUREG/CR-5561 (Reference 4). The bellows can withstand large pressure loading but may tear once the containment vessel deflection becomes large. Testing reported in NUREG/CR-6154 (Reference 26) has shown that the bellows remain leaktight even when j
subjected to large deflections sufficient to fully compress the bellows. Such large deflections do not occur as long as the containment vessel remains elastic. As described in subsection 3.8.2.4.2.6, the radial deflection of the shell increases substantially once the containment cylinder yields. The resulting deflections are assumed to cause loss of containment function.
The containment penetration bellows are designed for a pressure of 90 psig at design temperature within Service Level C limits, concurrent with the relative displacements imposed j
on the bellows when the containment vessel is pressurized to these magnitudes.
Electdcal penetrations have a pressure boundary consisting of the sleeve and an end plate containing a series of modules. The pressu~ capacity of these elements is large and is greater than the capacity of the containment vesse..ylinder at temperatures up to the containment design temperature. Electrical penetration assemblies are also designed to satisfy ASME Service Level C stress limits under a pressure of 90 psig at design temperature. Tests at pressures and tempe.atures representative of severe accident conditions are described in NUREG/CR-5334 (Reference 5), where the Westinghouse penetrations were irradiated, aged, then tested to 75 psia at 400*F. Other electrical penetration assemblies were tested to higher pressures and temperatures. These tests showed that the electrical penetration assemblies withstand severe accident conditions. The electrical penetration assemblies are qualified for the conta! ament design basis event conditions as described in Appendix 3D. The assemblies are similar to one of those tested by Sandia as reported in NUREG/CR 5334 (Reference 5).
He ultimate pressure capacity of the electrical penetration assemblies is primarily determined by the temperature. The maximum temperature of the containment vessel below the operating deck during a severe accident is approximately equal to the containment design temperature of 280*F. This temperature is significantly below the temperature at which the assemblies from the three suppliers in the Sandia tests were tested.
3.8.2.4.2.6 Material Properties ne containment vessel is designed using SA537, Class 2 material. This has a specified minimum yield of 60 ksi and ultimate of 80 ksi. Test data for materials meeting SA537 or l
having similar chemical properties were reviewed. In a sample of 122 tests for thicknesses
[
7 Revision: 11 February 28,' 1997 3.8-14 T Westingt10ljC0
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Westinghouse Energy Systems Ba 355 Pittsburgh Pennsylvansa 15230 0355 Electric Corporation NSD-NRC-97-4981 i
DCP/NRC0737
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Docket No.: ST N-52-003 L
February 11, 1997 Document Control Desi.
i U. S. Nuclear Regulabry Commission Washington, DC 20555 1
ATTENTION T. R. QUAY i
SUBJECT:
AP600 CONTAINMENT OPEN ITEMS
Dear Mr. Quay:
l i
Attached are responses to resolve DSER open items, requests for additional information, and a telecon item related to AP600 containment structural issues. The responses include draft SSAR changes that '
l will be included in Revision 11 of the AP600 SSAR. The list below includes the DSER item number or RAI number and the open item tracking system (OITS) number.
1 DSER 013.8.2.4-3 (681)
DSER 013.8.2.4-20 (698)
DSER 013.8.2.4-28 (706)
DSER COL 3.8.2.4-1 (1888)
Telecon item June 23,1995 (2515) l RAI 220.100 (3269) l-RAI 220.101 (3270)
RAI 220.102 (3271)
~ These responses provide a way to resolve these items and will permit the NRC staff to provide input for the final safety evaluation report.
l If you have any questions please contact Donald A. Lindgren at (412) 374-4856.
lPM L
Brian A. McIntyre, Manager
- Advanced Plant Safety and Licensing
/jml
~
Attachment 77 i
cc:
D. Jackson, NRC L
T. Cheng, NRC
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1 l 5 l
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Attachment to NSD-NRC-97-4981 l
l Open item #706 - DSER Open Item # 3.8.2.4-28 Westinghouse should provide in the SSAR sn assessment of the pressure capability of the main steamline and main feedwater line bellows, a corresponding failure probability distribution curve, and the impact on the overall cumulative failure probability curve.
Westinghouse response This is the same question as RAI 480.191. See response to RAI 480.191 provided in letter j
NSD-NRC-96-4904, dated 12/9/96.
1 l
Open Item #1888 - DSER Open item # 3.8.2.4-1 The COL applicant should demonstrate that EPAs to be used shall be at least as strong as the AP600 SCV.
Westinghouse resoonse Additional information has been included in SSAR subsection 3.8.2.4.2.5, Revision 7.
See also response to RAI 220.102 (Ol# 3271),
a Open Item #2515 - Telecon June 23,1995 Westinghouse should address the issue of fatigue and corrosion of containment bellows. The number of thermal cycles and loading information included in the design specification should be addressed.
The material requirements and effect of corrosion should also be included.
Westinghouse resoonse SSAR section 3.8.2.1.5, Rev 7 has been revind to include material and additional information on the displacement cycles. Fatigue is eva'uated in accordance with ASME subsection NE as stated in SSAR subsection 3.8.2.1.5. Bellows materials are stainless steel or nickel alloy. Corrosion is not expected; if there is any degradation it would be observed by inservice inspection or testing. The bellows are included in the ISI of the containment vessel as well as the containment leak rate testing.
1 i
DSER Open Item # 3269 (NRC letter dated 4/4/95) RAI # 220.100 In SSAR Section 3.8.2.4.2.3, the factor of safety (FS) of 1.67 is usad for equipment hatch covers ASME Service Level C limits.
Westinghouse estimated the critical buckling pressures for equipment hatches as 1.45 MPa (196 psig) for a 6.7 m (22 ft) diameter hatch and 1.21 MPa (161 psig) for a 4.9 m (16 ft) diameter hatch based on the classical buckling capacity of spherical shells subjected to external pressure and the capacity reduction factors specified in Baker et al., " Structural Analysis of Shells,"pp. 253-254, McGraw-Hill,1972, and in ASME Code Case N-284. The corresponding ASME Service Level C limits are 908 kPa (117 psig) and 763.2 kPa (96 p ig) using the factor of safety (FS) of 1.67 as specified in Code Case N 284, respectively. Q 3
x.
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Attachment to NSD-NRC-97-4981 Open Item # 3270 - RAI # 220.101 (NRC letter dated 4/4/96)
Westinghouse evaluated an additional BOSOR-5 analysis with stress-strain curves accounting for the effects of residual stresses on the buckling of cylindrical shells due to axial compression and/or external pressure. The failure mode was found to be an axisymmetric plastic collapse resulting from excessive vertical displacements at the pole. The maximum displacement was 1.09 m (43 in) at 1.45 MPa (195 psig). This information was requested by the staff to be provided in SSAR as discussed in RAI 6 (NRC letter dated September 14. 1995). In its response (NTD-NRC-96-4617 dated January 4, 1996), Westinghouse stated that the plastic collapse is bounded by the case for knuckle buckling without specific information. Provide this information in the SSAR.
Westinyhouse Resoonse Revised the second paragraph of SSAR subsection 3.8.2.4.2.2 as follows The top head was analyzed using the BOSOR-5 computer code (Reference 1). This code permits consider;, tion of both large displacements and nonlinear material properties. It calculates shell stresses and checks stability at each load step. Yield of the cylinder started at a pressure of 144 psig using elastic - perfectly plastic material properties, a yield stress of 60 ksi, and the von Mises yie!d criterion. Yield of the top of the crown started at an internal pressure of 146 psig. Yield of the knuckle region started at 152 psig. A theoretical plastic buckling pressure of 174 psig was determined. At this pressure, the. maximum effective prebuckling strain was 0.23 percent in the knuckle region where buckling occurred and 2.5 percent at the crown. The maximum deflection at the crown was 15.9 inches. A similar analysis was performed using nonlinear material properties considering the effects of residual stresses; buckling did not occur in this analysis, and failure would I occur once strains at the crown reach ultimate. The failure mode was found to be an axisymmetric I plastic collapse resulting from excessive vertical displacements at the crown. The maximum I displacement was 43 inches at 195 psig.
Open Item # 3271 - RAI # 220.102 (NRC letter dated 4/4/96)
In SSAR Section 3.8.2.4.2.5, mechanical and electrical penetrations are designed for a pressure of 90 psig at design temperature (280 F) for ASME Service Level C limits. In SSAR Section 3.8.2.4.2.8, however, the ASME Service Level C limit is 92 psig at 280 F from the containment ellipsoidal head plastic buckling. Clarify which pressure represents the ASME Service Level C limit at design I
temperature for the containment.
Westinghouse resoonse f
The SSAR revision proposed in the response to Open Item # 3269 shows pressures corresponding to l
the ASME Service Level C limit at design temperature for the containment based on three different acceptat.ce criteria. When the capacity is evaluated against ASME Service Level C limits, including stability limits with a factor of safety of 2.5, the pressure capacity at design temperature is 62 psig and is limited by the 16' diameter equipment hatch as shown in proposed SSAR Table 3.8.2-2.
- p. f m
8
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