ML20141K189
| ML20141K189 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 05/15/1997 |
| From: | Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD-NRC-97-5131, NUDOCS 9705280359 | |
| Download: ML20141K189 (227) | |
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Westinghouse Energy Systems Nx355 Electric Corporation Pittsburgh Pennsylvania 15230-0355 NSD-NRC-97-5131 DCP/NRC0867 Docket No.: STN-52-003 May 15,1997 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 March 14,1997 through April 11, 1997.
Attachment 1 provides the index of the attached material as you have requested.
A Brian A. McInt).., h a tag r Advanced Plant Safety and Licensing jml Attachment Enclosure cc: N. J. Liparulo, Westinghouse (w/o Attachment, Enclosure) _
M. M. Slosson, NRC (w/o Enclosure)
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9705280359 970515 PDR ADOCK 05200003 {j\hl\hl[\}$\}l[hkllIiklk l A PDR . . a * *
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Attachment I to Westinghouse Letter DCP/NRC0867 DATE ADDRESSEE DESCRIPTION 4/2/97 Kenyon Open items for Chapter 12 as discussed in 4/1 phone call. Will be in Revision 12 unless we hear otherwise.
3/14/97 Quay 3/7 fire protection meeting summary.
3/24/97 Jackson Revision of 3/21 SSAR markup to resolve open items 472 and 1172. Will be in Revision 12 unless we hear othe; wise.
3/27/97 Quay /Scaletti Open items status summacy.
4/2/97 Jackson VAS design change summar". Will be in Revision 12 unless we hear otherwise.
4/2/97 Scaletti information related to open item 213. Was submitted in revision 7 (5/6/96). Request NRC review material and provide definitive action or provide direction to change NRC status to Action N or closed.
4/2/97 Scaletti Information related to open item 214. Was submitted in revision 7 (5/6/96). Request NRC review material and provide definitive action or provide direction to change NRC status to Action N or closed.
4/2/97 Scaletti infor.1ation related to open item 3. Material submitted 2/21/97. Request NRC review material and provide definitive action or provide direction to change NRC status to Action N or closed.
4/2/97 Scaletti Information related to open item 210. Was submitted in revision 7 (5/6/96). Request NRC review mater;d and provide definitive action or provide direction to change NRC status to Action N or closed.
4/2/97 Scaletti information related to open item 212. Was submitted in revision 7 (5/6/96). Request NRC review material and provide definitive action or provide direction to change NRC status to Action N or closed.
4/2/97 Scaletti information related to open item 219. Was submitted m I revision 7 (5/6/96). Request NRC review material and provide !
l definitive action or provide direction to change NRC status to Action N or closed. l 4/3/97 Scaletti Information related to open item 210. Was submitted in revision 7 (5/6/96). Request NRC review material and provide l definitive action or provide direction to change NRC status to Actien N or closed.
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4/3/97 Scaletti information related to open item 222. Was submitted in i revision 8 (7/96). Request NRC review material and provide l definitive action or provide direction to change NRC status to l Action N or closed.
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! 4/3/97 Scaletti information related to open item 216. Was submitted in l revision 8 (7/96). Request NRC review material and provide l definitive action or provide direction to change NRC status to Action N or closed.
4/3/97 Scaletti Information related to open item 153. Was submitted in fax on 3/10/97. Request NRC review material and provide definitive action or provide direction to change NRC status to Action N or closed.
4/3/97 Scaletti information related to open item 213. Was submitted in revision 7 (5/6/96). Request NRC review material and provide definitive action or provide direction to change NRC status to Action N or closed.
4/4/97 Scaletti Information related to open item 4. Was submitted in letter dated 3/13/97. Request NRC review material and provide definitive action or provide direction to change NRC status to Action N or closed.
4/4/97 Scaletti Information related to open item 207. Was submitted in revision 7 (5/6/96). Request NRC review material and provide definitive action or provide direction to change NRC status to l Action N or closed. l 3/27/97 Jackson Markup of SSAR 3.9. Returns information previously struck i out. Will be included in SSAR revision 12. l I
3/27/97 Jackson Markup changes to SSAR section 3.11 and Appendix 3D as !
discussed in cailier phone call.. Will be included in Revision 12.
4/1/97 Sebrosky Page change that will be included in Revision 9 to the PRA.
Figure 44-1.
4/1/97 Sebrosky Request to acknowledge receipt of information related to numerous PRA open items. Request NRC review material and provide definitive action or provide direction to change NRC status.
3/31/97 Sebrosky Notice that open items 1429,1430 and 1431 should be closed based on Revision 8 to the PRA. Request NRC review material and provide definitive action or provide direction to change NRC status.
l 4/2/97 liuffman Request for NRC to provide status on open item 3137.
l 4/2/97 11uffman Update of tech spec for 4/2 phone call.
( 4/2/97 Jackson Comments on draft 3/5 WGOTillC meeting notes.
32(mA
. w n -o 3/31/97 Iluffman Information for numerous NOTRUMP related items for which Westinghouse believes the status should be changed to closed.
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1 4/3/97 Quay /Scaletti Open item status summary.
4/10/97 Sebrosky SSAR markups to address I&C ITAAC comments during 4/1/7 ;
phone call. '
4/9/97 Jackson SSAR markup to resolve reviewer concern on Class D vs.
Class E. Will go into revision 12 unless we hear otherwise.
l 4/8/97 Quay /Iluffman Summary of 4/3 RTNSS meeting.
4/6/97 Quay Request for copy of 1/7/97 letter mentioned in Martin letter of l 3/27/97.
4/7/97 Iluffman Information related to DSER open CN-21.6.2.4-4 for 4/7 phone I call.
4/7/97 Iluffman Explanation of why grid should be stable for at least 3 seconds.
4/3/97 Iluffman Draft markups for Chapter 8 and table 1.8 of SSAR to reflect 3 second delay. Preparation of upcoming phone call.
4/11/97 iluffman Draft tech spec 3.7.8.
4/9/97 Scaletti information related to open item 267. Was submitted in revision 7 (5/6/96). Request NRC review material and provide definitive action or provide direction to change NRC status to Action N or closed. .
4/7/97 Scaletti information related to open item 207. Was submitted in revision 7 (5/6/96). Request NRC review material and provide definitive action or provide direction to change NRC status to 1 Action N or closed. 1 4/7/97 Scaletti information related to open item 274. Was submitted in revision 7 (5/6/96). Request NRC review material and provide definitive action or provide direction to change NRC status to Action N or closed.
1 4/7/97 Scaletti information related to open item 297. Was submitted m '
revision 7 (5/6/96). Request NRC review material and provide definitive action or provide direction to change NRC status to Action N or closed.
4/8/97 Scaletti information related to open item 269. Was submitted in j revision 7 (5/6/96). Request NRC review material and provide '
definitive action or provide direction to change NRC status to Action N or closed. l 4/8/97 Scaletti information related to open item 271. Was submitted in revision 7 (5/6/96). Request NRC review material and provide definitive action or provide direction to change NRC status to Action N or closed.
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FAX TO TOM KENYON l r
i l To: E. Cummins l l Ron Vijuk B. McIntyre
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l J. Sejvar l l T. Hayes i l G. Israelson -
j Jeanne Evans ,
Attached are the SSAR markups for 4 of the remaining 5 open items on Chapter 12. They should be :
i as we discussed them in'a telephone conversation on April 1,1997. They will go into Revision 12 l unless we hear from you. As we discussed, the fifth item will be submitted next week after we have !
I had some time to investigate the resolution path.
4 Jim Winters l l !
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s 11. Radioactive Waste M=giment input from this flow sensor. This signal is used by the radiation processor to control sample flow. The analog signal is transmitted to the plant control system (protection and safety monitoring system for safety-related monitors). For offline liquid monitors, a flow indicator is provided for manual adjustment of the flow.
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11.5.2.3.1 Fluid Process Monitors '
Steam Generator Blowdown Radiation Monitors The steam generator blowdown radiation monitors (BDS-JE-RE010, RE011) measure the l concentration of radioactive material in the blowdown from the steam generators. One )
measures radiation in the purification process effluent before it is retumed to the condensate system. The other measures radioactivity in the blowdown system electrodeionization waste brine before it is discharged to the waste water system. 'Ihe presence of radioactive material in the steam generator blowdown indicates a leak between the primary side and the secondary )
side of the steam generator. Refer to subsection 5.2.5 for details of leakage monitoring and ,
to subsections 10.4.8 and 11.2 for process system details. The steam generator blowdown '
radiation monitors meet the guidelines of Regulatory Guide 1.97 as discussed in Appendix 1 A and Section 7.5.
l AP600 has two steam generators, each of which has a blowdown line. Each blowdown line i has a heat exchanger upstream of the blowdown flow control valve. The steam generator i blowdown radiation detectors are located in the lines downstream of these heat exchangers. j Therefore, the radiation monitors do not require a sample cooler.
When its predetermined setpoint is exceeded, each steam generator blowdown radiation monitor initiates an alarm in the main control room, initiates closure of the steam generator blowdown containment isoladon valves and the steam generator blowdown flow control valves, and diverts flow to the liquid radwaste system.
The steam generator blowdown radiation moniton we inline gamma sensitive, thallium-activated, sodium iodide scintillation detectors. The steam generator blowdown radiation monitor detector range and principal isotopes are listed in Table 11.5-1.
The arrangement for the steam generator blowdown radiation monitor is shown in Figure 11.5-1.
Component Cooling Water System Radiation Monitor The component cooling water system radiation monitor (CCS-JE-RE001) measures the concentration of radioactive material in the component cooling water system. Radioactive material in the component cooling water system provides indication of leakage. Refer to subsection 5.2.5 for details of leakage monitoring and to subsection 9.2.2 for process system details.
Revision: 9 2&7 August 9,1996 11.5-4 [ W65tingh0USS J
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f .j 12. R di tion Protection l L _ __
Clean Services l
Whenever practicable, clean services and equipment such as compressed air piping, clean ;
water piping, ventilation ducts, and cable trays are not routed through radioactive pipeways. I hiaterials l
Equipment specifications for components exposed to high temperature reactor coolant contain l i limitations on the cobalt content of the base metal as given in Table 12.3-1. The use of hard facing material with cobalt content such as stellite is limited to applications where its use is I
necessary for reliability considerations. Nickel-based alloys in the reactor coolant system (Co-58 is produced from activation of Ni-58) are similarly used only where component reliability may be compromised by the use of other materials. He major use of nickel-based 1 I '
alloys in the reactor coolant system is the inconel steam generator tubes.
gawer a. a. i. l. I &
Single Integrated Gripper hiast Assembly Refueling Afachine l
To minimize the radiation exposure during refueling, a smgle integrated gripper mast assembly refueling machine is used. The machine permits removal and insertion of thimble plugs or rod control cluster assemblies while a fuel assembly is being handled by the refueling machine.
1 Improved IIead Closure System '
The head closure system is designed to minimize the reactor head stud tensioning time.
12.3.1.1.2 Common Facility and Layout Designs for ALARA I
This subsection describes the design features utilized for standard plant process and layout situations. Rese features are employed in conjunction with the general equipment described in subsection 12.3.1.1.1 and include the features described in the following paragraphs.
Valve biodules l Selected valve modules are provided with shielded entrances for personnel protection. Floor drains are provided to control radioactive leakage. To facilitate decontamination, concrete i surfaces are covered with a smooth surface coating which allows decontamination.
Piping Pipes canying radioactive materials are routed through controlled access areas properly zoned I
for that level of activity. Radioactive piping runs are analyzed to determine the potential radioactivity level and surface dose rate. Where it is necessary that radioactive piping be I
routed through corridors or other low radiation zone areas, shielded pipeways or distance i
separation are provided. Whenever practicable, valves and instruments are not placed in Revision: 7 3S7 i April 30,1996 12.3-6 T Westinghouse 4
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- 12. Radiation Protection i 12.3.2.2.9 Spent Fuel Transfer Canal and Tube Shielding i The spent fuel transfer tube is shielded to within adjacent area radiation zone limits. This is primarily achieved through the use of concrete and water. The only removable shielding consists of concrete or s' eel hatches which reduce radiation in accessible areas to within those levels prescribed in the normal operation radiation zone maps (Figure 12.3-1).
The spent fuel transfer tube is completely enclosed in concrete and there is no unshielded I portion of the spent fuel transfer tube during the refueling operation. The only potential I radiatien streaming path associated with the tube shielding configuration is the 2 inch I (5.08 cm) seismic gap between the fuel transfer tube shielding and the steel containment wall.
I Shielding of this gap is provided by a water-filled bladder. This " expansion gap" radiation I shield provides effective reduction of the radiation fields during fuel transfer and I
accommodates relative movement between the containment and the concrete transfer tube I
shielding with no loss in shield integrity. A removable hatch in the shield conf pration provides access for inspecti_on of the fuel transfer tube welds. The opening of :his hatch is administratively controll is hatch is in place during the snent fuel transfer operation.
Iasewrxrevsetr12.L2.2.Q 12.3.2.3 Shielding Calculational Methods The shielding thicknesses provided for compliance with plant radiation zoning and to minimize plant personnel exposure are based on maxiraum equipment activities under the plant operating conditions described in Chapter 11 and Section 12.2. The thickness of each shield wall surrounding radioactive equipment is determined by approximating as closely as practicable the actual geometry and physical condition of the source or sources. The isotopic concentrations are convened to energy group sources using data from standard references (References I through 6).
The geometric model assumed for shielding evaluation of tanks, heat exchangers, filters, ion exchangers, and the containment is a finite cylindrical volume source. For shielding evaluation of piping, the geometric model is a finite shielded cylinder. In cases where radioactive materials are deposited on surfaces such as pipe, the latter is treated as an annular cylindrical surface source.
The computer code SHIELD-SG (Reference 11) is used to calculate dose rates. For complex
! geometries other computer codes such as QAD (Reference 16) are used. Buildup, calculated using Berger coefficients presented in ORNL-RSIC-10 (Reference 7) and Blizard's Method-of Buildup Determination, presented in the Engineering Compendium on Radiation Shielding I (Reference 8), is used for laminated shields.
The source activity (Ci) and gamma ray source strengths (MeV/sec) are calculated using one of the following computer codes: ORIGEN (Reference 17), SOURCE 2/ACCUM (Reference 12), or RADGAS3 (Reference 13). ACCUM (Reference 12) is an option wi W l SOURCE 2 that computes isotope accumulation for several time periods from a given flow of isotopes in curies per second. His accumulated activity may then be decayed for any number of decay times at which gamma energy spectra and isotope Curie activity an: computed. He Revision: 7 "/4 7 April 30,1996 12.3-14 T Westinghouse J
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- 12. Rrdlation Protection i
l of advanced technology into the refueling process also reduces doses. Table 12.4-11 lists some of the AP600 features that reduce doses during refueling operations.
l Table 12.4-12 provides dose estimates for the various refueling activities.
12.4.1.7 Overall Plant Doses The estimated annual personnel doses associated with the six activity categories discussed above are summarized below:
Estimated Annual Category Percent of Total Dose (man. rem)
Reactor operations and surveillance 20.6 13.8 Routine inspection and maintenance 18.0 12.1 Inservice inspection 24.6 16.5 Special maintenance 22.4 15.0 Waste processing 7.8 5.2 Refueling 6.6 4.4 Total 100.0 67.0 These dose estimr.tes are based on operation with an 18-month fuel cycle and are bounding for operation with a 24-month fuel cycle.
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12.4.2 Radiation Exposure at the Site Boundary 12.4.2.1 Direct Radiation The direct radiation from the containment and other plant buildings is negligible. The AP600 design also provides storage of refueling water inside the containment instead of in an outside storage tank that eliminates it as a radiation source.
12.4.2.2 Doses due to Airborne Radioactivity Subsection 11.3.3 discusses doses at the site boundary due to activity released as a result of normal operations.
12.4.3 Combined License Information This section has no requirement for information to be provided in support of the Combined License application.
Revision: 9 S ") 7 August 9,1996 12.4-4 [ WBSilingh0US8
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- Insert 11.5.2.3 i
Those airborne radiation monitors which monitor plant areas which may be occupied by plant personnel will be capable of detecting 10 DAC-hours. The specific radiation monitors which are included in this category are identified in Table 11.5-1.
Insert 12.3.1.1.1 General prohibitions on antimony and other low melting point metals are contained in subsection 6.1.1. In addition. the reactor coolant pump mechanical design criteria prohibits antimor.y completely from the reactor coolant pump and its bearings.
l Insert 12.3.2.2.9 l
,c and is treated as an entrance to a very high radiation area under 10CFR20. T Insert 12.4
- 12. L t.8 Post Accident Actions l
Requirements of 10 CFR 52.79(b) relative to plant area access and post-accident I sampling (10 CFR 50.34 Item (2)(viii)) are included in Section 1.9.3. If procedures i are followed. the design limi ts radiation exposures to any individual to not exceed 5 i rem to the whole-body or 75 rem to the extremities. P!=: == fer p=: =!d:at l p=c=:! === = :ddr=d 5 S=:!cn 12.3, :=!uding $: =dia:!ca == =p: I
- =!ud:d = Fig = 12.3 2. Figure 12.3-2 in Section 12.3 containl radiation zone; maps .
for plant areas including those areas requiring post accident access { This figure shows i
projected radiation zones in areas requiring access and access routes for ingress,
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egress and performance of actions at these locations. The radiation zone maps reflect 1 maximum radiation fields over the course of an accident. The .nalyses that confirm that the d=: P-!:: individual personnel exposure limits following and accident are not exceeded reflect the time-dependency of the area dose rates and the required !
post-accident access times. The areas that require post-accident accessibility are:
- 1) Main control room
- 2) Primary sampling room
- 3) Class IE regulating transformer areas
- 4) Ventilation control area for I&C rooms with PAMS equipment
- 5) Valve area to align spent fuel pool makeup
- 6) Ancilla'y diesel room
- 7) Passive containment cooling water inventory make-up area The area which results in the highest individual personnel exposures is the primary sampling room. The design provides for access to the primary sampling room as early as eight hours after the accident when radiation fields are high compared to 64 hours7.407407e-4 days <br />0.0178 hours <br />1.058201e-4 weeks <br />2.4352e-5 months <br /> or later for the other areas requiring access outside the main control room. in addition to the design provisions, individual exposure for this early sampling operation may be minimized by proper administrative operational controls (for example, splitting tasks'among different crews or limiting sample sizes). Special operational 6,47
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} controls would only be considered in the event that radiation fields, associated with l access to the primary sampling room reach the conservatively high levels considered '
i in the evaluations. These conservatively high levels include activity releases as l defined in NUREG-1465, maximum design basis leak rate from containment into the l
' access areas and no operable building ventilation systems.
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- TX CONFIRMATION REPORT ** AS OF APR * '97 10:19 PAGE.01 1
AP600 DESIGN CERT DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 4/ 2 10:15 #23:NRC )
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B, rian A. McIntyre,06:12 PM 3/14/97,35 fire protection meeting _ _ _ _ _ _ __
Date: Fri,14 Mar 199718:12:48 -0500 I To: TRQ@NRC. GOV From: " Brian A. McIntyre" <meintyba@wesmail.com> - ,
Subject:
3R fire protection meeting Cc: meintyba@wesmail.com l
- Ted, This is anothe tview of the fire protection meeting from an internal report.
We are growing increasingly concerned that this area will not be closed out in a timely manner given the progress that has been made to date. Our observation is that additional resources are needed at the reviewer level. I This is already a " Top 27" issue and will be on the agenda for the next !
senior management meeting. Lets try the week of April 7 for that meeting. i I will check to see if Howard, Bob and Ed are available and give you a call.
It would be helpful if we had the position letters that Tim indicated were in the works at the March 3 meeting in time to discuss them at the next meeting. l l
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"Our March 7,1997 meeting with NRC on fire protection was not as productive as expected. Westinghouse came prepared to explain its position on " cold" versus " safe" shutdown for fire protection, protection of fire main water supply in the event of a turbine hall fire, and smoke control logic. It was not apparent to us that the NRC had performed any review, coordination, or their action items since the last meeting on February 10,1997. No items were closed, but Westinghouse did provide its position to NRC management (section lead Steve West). Mr. West agreed that it is time for NRC to review what Westinghouse had provided. Westinghouse must formally document the information provided at the meeting. This will be complete this week."
Brian A. McIntyre 412.374.4334 WIN 284.4334 FAX 374.4887
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Printed 'for " Brian A. McIntyre" <m'cintyba@wesmail.com> [ ~~ 1[
W Westinghouse FAX COVER SHEET RECIPIENT INFORMATION SENDER INFORMATION DATE: ,dAnu 24 /99 7 NAME: L Q,gy TO: LOCATION:
ENERGY CENTER -
/ (4N6- d Acic.snev EAST PHONE: FACSIMILE: PHONE:
Office:t/I L-3 7'/-529o COMPANY: Facsimite: Win: 284 4887 dIA outside: (412)374 4887 LOCATION:
Cover + Pages 1+/
The fo!!owing pages are being sent from the Westinghouse Energy Center, East Tower, 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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Coteria 1(ef(renced AP600 Section. - Criteria Position ClarifictiorvSummary Dcscripdon of Exceptions C.I.l.3 Conforms The auxiliary building at contains the liquid radwaste system is designed to Seismic Category I criteria. The Seismic Category I nructure will retain the maximum liquid gb- inventory of the IW. The lowest level of the auxiliary building, cicration 66'6". contains the liquid radwaste system effluent holdup tanks, waste holdup tanks, a monitor tank and chemical waste tank within a common flood zone. This flood zone has watertight floors and walls. The enclosed solume within this flood zone is
_s_ufficient to contain the contents of the tanke SY y l Within the Good zone, theTffliitmtttdup tanks, Iw holdup tanks, chemical waste tank, and monito nk are each in a separate gr, which is accessed
' climbing downJs Hifer. The only point in the a iliaryjb R6ing with a lower i elevation is the au Mary building sump.flhe Et ank rooms -
l ave o or two Iloor ora ns that g f lead to sump. Tank ov ows or spills will be c ected in the auxiliary bui sump. The f sump is automatically pumped to a w holdup L na 4 e -
bTwo liquid radwaste system monitor tanks are three levels up at elevation 100"0". Overflows or 'Mj8 from these monitor tanks drain by gravity down through the drain system to a waste holdup tank.
b ' or tanks are each in separate r s, which are acces bin a ladder.
The tank rooms e one r drain thab leads to a aste holdup tank. Qe Seismic fili Category I criteria exceed the operating basis "
carthquake required by regulatory position C.5 of Regulatory Guide 1.143.
C.I.l.4 Conforms Components in the liquid radwaste systems are non<,cismic, lhey are not required to be designed for seistrde loads.
C.I.2.1 Conforms Atmospheric tanks in the liquid radwaste system have level senso.s, transmitters, and alarms.
Local alarm is not provided because the tanks are located in shielded areas that are not normally occupied by people.
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FAX TO DIANE JACKSON April 2,1997 cc: Hutchings Cummins R. M. Vijuk McIntyre Lindgren Jeanne Evans Here is a package of information related to the design change I discussed with you last week. It is being sent for your preliminary review. It concerns a modification to the VAS. In order to simplify arrangements and to have a simpler, more maintainable, and more cost effective system, we combined the radioactive chemistry laboratory HVAC subsystem with another subsystem within VAS. This change will go into Revision 12 of the SSAR unless we hear from you that it will cause a delay in FSER section 9.4 or if it is unacceptable to the reviewer. I'll call you soon to set up a telephone call to further explain this package, if required.
Thanks Jim Winters 412-374-5290 l
DCP GW-GEE-447/0P
- Page 5 Description of Design Change Proposal for Deletion of the Rad Chem Lab HVAC Subsystem Backcrouni The radiologically controlled area ventilation system (VAS) serves the radiation chemistry laboratory, primary sample room and security rooms located within the radiologically controlled areas of the auxiliary building. These areas are served by a separate supply air subsystem consisting of two 100 percent capacity supply air handling units and a single electric humidifier located in the common ductwork downstream of the supply air handling units. Each supply air handling unit has a HEPA filter bank, a hot water heating coil and a supply fan designed to maintain ambient room temperatures between 73-78 F with a relative humidity of 35-50% in accordance with URD Chapter 9 paragraph 8.2.6. Because these areas are considered to be normally occupied,100 percent redundant units are provided in accordance with URD Change Notice 445.
The URD requires HEPA filtration of supply air based on previous PWR design considerations to support calibration of equipment sensitive to airborne particulate (nonradioactive) contamination. AP600 does not utilize equipment that requires HEPA filtration of supply air. In a meeting with the USG on 11/21/96, it was agreed that a separate ventilation system with HEPA filtration is not necessary. (Ref: URD Finding
$9520) Design room temperatures, however, should not exceed 80 F assuming major equipment failure.
Desien Chance: #
Because HEPA filtration is not required, the dedicated supply air handling units (VASp S-003A/B) which are located in the annex building can be replaced with local duct electric reheat coils and humidifierslocated in the auxiliary building, ne radiation chemistry laboratory will be realigned to the supply and exhaust air ducts branches that serve the auxiliary building to eliminate the need to install lorig runs of dedicated HVAC ductwork between the auxiliary and annex buildings. Additionally, the primary sample room is temporarily occupied only when fluid samples are being collected. Its design room te(mperature range will be revised from 73-78'F to 50-1(M*F based on the URD criteria for plant areas with? frequent inspections.
Humidification of supply air for the primary sample room will be eliminated he supply airflow rate to the radiation chemistry laboratory and security rooms will be increased so that when one 50% capacity aux / annex building supply air handling unit is shutdown for maintenance, there will be adequate flow to maintain design room temperature conditions in accordance wit URD Change Notice 445. Herefore, continuous reheat is required so that these rooms are not over-cooled during normal plant operation.
Electric reheat coils will be provided because the hot water heating system (VYS) is not intended to be operated on a continuous basis. Each reheat coil will be furnished with 2 stages of electric heat to provide backup heating elements in the event there is a failure of one heating stage. He rad chem lab pressure differential controls used to modulate the supply fan inlet vanes will be deleted because they are not required to compensate for changes in the HEPA filter loading. The supply and exhaust airflow will be manually balanced to maintain a slightly negative air pressure with respect to adjacent access corridors.
YY
J . r 1 DCP GW-GEE-447/0P l: Page 6 s
Justification:
i This change does not affect safety-related or defense-in-depth equipment, It maintains compliance with the URD design room temperature, pressure and equipment redundancy criteria while reducing the total number of fans, dampers, nlters and quantity of ductwork installed in the plant.
Eauioment Sirine:
Eauioment Number Eauioment Descrintion Electrical Reauirements DWS Reauirements VAS EH 01 Electric Duct Heater 25 Kw/ stage -
VAS EH 02 Electric Duct Heater 10 Kw/ stage -
VAS MY H01 Humidifier 30 Kw 0.3 gpm VAS MY H02 Humidifier 15 Kw 0.2 gpm SSAR Changes: ,
See Pages 8 through 12 for markup.
P&lD Changes:
See.Pages 13 through 17 for markup. >
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- 9. Auxiliary Systams kojd 8 5 l
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i Spent resin equipment rooms
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. 50 - 130 I Radioactive pipe chases and valve rooms .
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..... .... 73 - 78 _
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....... .. . .. . . . .. . . . .. 73 - 78 1
9.43.2 System Description I
ne radiologically controlled area ventilation system consists of the following subsystems:
I I
- Auxiliary / annex building ventilation subsystem '
l .
Full handling area ventilation su,bs.y. stem s __
1 try laboratory l
I ne defense in depth portion of the system is sh in_ Figure 9.4g __
$j//y ena'sforete 0/wt., conhinmem 9.4.3.2.1 General Description pu gj,,gf gjf,4,7,,,
I 9.4.3.2.1.1 Auxiliary / Annex Building Ventil Qccess Cor$ der; on/, gegy,r jy, n Subsystem -
l Oncb \
l The auxiliary / annex building en ~ ation subsystem serves radiologically co_ntrolled equi ment, ,
I piping and valve room [ adjacent access and staging areag~thelradiation _chermstry I
atsoratory ventlauon subsyst'edt. He auxiliary / annex building venulation subsysfem consists I
of two 50 percent capacity supply air handling units, a ducted supply and exhaust air system, I isolation dampers, diffusers and registers, exhaust fans, automatic controls and accessories.
I The supply air handling units are located in the south air handling equipment room of the I
annex building at elevation 158'-0". ne units discharge into a ducted supply distribution i system which is routed through the radiologically controlled areas of the auxiliary and annex l buildings. He supply and exhaust ducts have isolation dampers that close to isolate the I
auxiliary and annex buildings from the outside environment when high airbome radioactivity I is detected in the exhaust air duct. He supply and exhaust ducts _are configured so that two I building zones mr.y be independently isolated. The annex buildin adjac_ent iiuxiliary b_uilding sta ring, equipment areas, and 3 roomsFrved by the radiation c try laboratorv ventilannn
! secut'th systerruare aligned to one zone. ,ne other zone includes)primanlyfaawaste equipment l rooms, pipe chases, and Meess corridors (located in the auxiliari buiTdirI2 A]
I radiation monitor is located in the exhaust air duct from each zone.
f I
ne exhaust air fans are located in the upper radiologically controlled area ventilation system i 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 .
M =- m the rod'Nhon chemffry /eborafory, primo y .snapk room,"
spnt Ar/pse/ cec /me wafsnpump andheef e>rchauses m/ Revision: 1
[ Westhghouse d 9.4 27 / pril 30,1996 )
norme/twia'ud beatremovalpump and' heat e>cluanpr rooms, \
CVS
- _ _ _ malevy pwup room, law annale, middle amuln and various
_ ~
g" c =u DCPt GW-GEE-4t7/bP pe9 g 9. Auxiiiary syst:
I the low radiation access areas into the radioactive equipment and piping rooms with a greater i
potential for airborne radioactivity. Additionally, the exhaust air ductwork is connected to the I
radioactive waste drain system (WRS) sump to maintain the sump atmosphere at a negative I
air pressure to prevent the exfiltration of potentially contaminated air into the surrounding I
area. De exhaust air ductwork is connected to the radwaste effluent holdup tanks to prevent I
the potential buildup of airbome radioactivity or hydrogen gas within these tanks. He i
exhaust fans discharge the exhaust air ir,to the plant vent for monitoring of offsite airbome I radiological releases.
l The ventilation airflow dilutes potential airborne contamination to maintain the concentration I
at the site boundary within 10 CFR 20 (Reference 21) allowable effluent concentration limits I
and the intemal room airbome concentrations within 10 CFR 20 occupational derived air I
concentration (DAC) limits during normal plant operation.
i Unit coolers are located in the normal residual heat removal system (RNS) and chemical and
! I volume control system (CVS) pump rooms because they have significant cooling loads on an l I intermittent basis when large equipment is operating. Each unit cooler is sized to
- I accommodate 100 percent of its corresponding pump cooling load. The unit coolers are j l provided with chilled water from redundant trains of ti;e central chilled water system (VWS) l .
low capacity subsystem. De normal residual heat removal pump room unit coolers have two i
cooling coils per unit cooler so that chilled water supplied by either train A or train B alone I
can support concurrent operation of both normal residual heat removal system pumps. De I
two chemical and volume control makeup pump room unit coolers are connected to redundant I
trains of the chilled water system; however, operation of either the train A or train B unit I
cooler alone maintains the common makeup pump room temperature conditions and supports I operation of either makeup pump.
I Hearing coils are located in the supply air ducts serving plant areas that require supplemental I
heating during periods of cold outside air temperature conditions:fElectric unit heaters provide
, I supplemental heating in the middle annulus.
L l
ne upper annulus is separated from the middle annulus area of the auxiliary building by a l
I concrete floor section and flexible seals that connects the building. De annulus seal provides a passive barrier during normal plant operation or when I
the auxiliary building is isolated, preventint the exf2]tration of unmoniwa )
I sirom_thef middle annulus to the environment. 17/e meQ/ho [/jeinirdy 46dra andferurl/y I
roon supply air ducts erar mth local 9.4.3.2.1.2 Fuel Handling Ares Ventilation Subsy dec.fde. co//s on/ anric///hu foma/nfq/h l t menkl co with/n the areas
. I sus a en h /wrar Mr.sanaSco . _ . . ..
De fuel handling area ventilation subsy sicTuel rtansmg area, car oaymner I
storage area, and the spent resin equipment and piping rooms. De fuel handling area I
ventilation subsystem consists of two 50 percent capacity supply air handing units, a ducted I
supply and exhaust air system, isolation dampers, diffusers, registers, exhaust fans, automatic i I
controls and accessories. The ventilation airflow capacity is designed to maintain I
environmental conditions that support worter efficiency during fuel handling operations based I
on a maximum wetbulb globe temperanut of 80*F (%'F drybulb) as defined by EPRI
(
Revision: 7 @
April 30,1996 9.4-28 Y WCStingh00$8 a
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' 1 DCP: GW-6EE-447/OP g
- 9. Auxiliary Systems Pup 10 i
i NP-4453 (Reference 22). De supply air handling units are located in the south air handling ;
I equipment room of the annex building at elevation 135' 3" The units discharge into a ducted i I
supply distribution system which is routed to the fuel handling and rail car bay / filter storage l
- I areas of the auxiliary building. The supply and exhaust ducts are provided with isolation l l dampers that close when high airborne radioactivity in the exhaust air or high pressure !
I differential with respect to the outside atmosphere is detected. l I
ne exhaust air fans are located in the upper radiologically controlled area ventilation system i
equipment room at elevation 145'-9" of the auxiliary building. The supply and exhaust I ductwork is arranged to exhaust the spent fuel pool plume and to provide directional airflow I
from the rail car bay / filter storage area into the spent resin equipment rooms. The exhaust i , fans discharge the exhaust air into the plant vent for monitoring of offsite airbome i radiological releases.
I ne ventilation airflow dilutes potential airborne contamination to maintain the concentration I at the site boundary within 10 CFR 20 (Reference 21) allowable effluent concentration limits '
I and the intemal room airbome concentrations within 10 CFR 20 occupational derived air I concentration (DAC) limits during normal plant operation.
I h4.3.2.1.3 Radiation Chemistry Laboratory Ventilation Subsystem { .
l The radiation chemistry laboratory ventilation subsystem serves the radiation ch I laboratory, primary sample room and auxiliary building security rooms. He radiation I
chemistry laboratory ventilation subsystem consists of two 100 percent capacity supply air { ,
I handing units, a ducted supply air system, a humidifier, diffusers, registers, automatic controls '
I and accessories. De supply air handling units are located in the south air handling equipment I '
room of the annex building at elevation 158'-0". De supply air handling units are connected I
to the auxiliary / annex building ventilation subsystem supply air duct to utilize preconditioned l
and prefiltered outdoor air. Supplemental filtration is provided by the radiation chemistry j l I laboratory ventilt. tion subsystem for added cleanliness to support operation of sensitive f I I equipment. A humidifier is located in the common supply air ducrwork downstream of the I supply air handling units. He radiation chemistry laboratory exhaust air is ducted to the I auxiliary / annex building ventilation subsystem exhaust fans. De ventilation airflow dilutes 1 I
room intemal airbome radioactivity concentrations within 10 CFR 20 occupational derived air, I
geentration (DAC) limitsf 9.4.3.2.2 Compoocat Description The radiologically controlled area ventilation system is comprised of the following major j l components. Equipment classified as Class A, B, C or D and applicable codes and standards l 1 are provided in Section 3.2. Table 9.4.3-1 provides design parameters for major defense in I
f depth components in the system.
4 l
4 g((] Revision: 7
[ W95fbgh00$8 9.4 29 April 30,1996 l
g DCP:G W-GEE-447/oP ^ ^"ri!!*'7 systuns
, . , , J Pa y lI Supply Air Handling Units Each supply air handling unit consists of a low efficiency filter bank, a high efficiency filter I
bank, a hot water heating coil bank, a chilled water cooling coil bank, and a supply fan,t e I
(radiation chemistry laboratory supply air handling units only consist of a liigh etnciency filter I
(bank, a hot water heating coil bank and a supply fanf I Supply and Exhaust Air Farts I
The supply and exhaust air fans are centrifugal type, single width single inlet (SWST) or I
double width double inlet (DWDI), with high efficiency wheels and backward inclined blades I
to produce non-overloading horsepower characteristics. The fans are designed and rated in I
accordance with ANSUAMCA 210 (Reference 4), ANSI /AMCA 211 (Reference 5), and i AMCA 300 (Reference 6).
I
- I Unit Coolers l
l Each unit cooler consist of a low efficiency filter bank, a chilled water cooling coil bank and I
a supply fan. The normal residual heat removal system pump room unit coolers have i redundant cooling coil banks.
l l Low and High Ef5ciency Filters i
I The low efficiency filters and high efficiency filters have a rated dust spot efficiency based i
on ASHRAE 52 (Reference 7). The filters minimum average dust spot efficiencies for the i
defense in depth filters are shown in Table 9.4.31. ne filters meet UL 900 (Reference 8) l Class I constnaction criteria.
I I Electric Unit Heaters 1
I The electric unit heaters are single stage or two-stage fin tubular type. De electric unit heater I
are UL-listed ud meet the requirements of UL 1025 (Reference 25) and National Electric i i Code. -
I kn. Heating Coils g,9 gy, 1 Thebting coils are(hot __ water finned tubular type. The outside supply air he i
provided with integral ~ face and bypass dampers to prevent freeze damage when modulating 1
the heat output. Coils are performance rated in accordance with ARI 410 (Reference 12).
t I
I
'M Cooling Coils l{
lI ne chilled water cooling coils are counterflow, finned tubular type. The cooling coils are i designed and rated in accordance with ASHRAE 33 (Reference 11) and ANSI /ARI 410 1
(Reference.& a
' ectric Heating Colls The electric heating coils are multi stage fin cubular type. De electric heating coils meet the Revision: 7 requiremenu of UL 1096 (Referene t y April 30,1996 ~9,4.Jo g igg,g
DCP GW-GEE-%7/oP Page 12 9. Auxiitary systems l
l
\
l occurring within the fuel handling area. fire dampers automatically isolate the HVAC i I
ductwork penetrating this fire area when the local air temperature exceeds predetermined l l serpoints.
! l 9.4.3.2.3.3 Radiation Chemistry Laboratory Ventilation Subsystem '
Normal Plant Operation i
During normal plant operation, one of two supply air handling units operates continuously to t i
ventilate the areas served on a once through basis. The supply airflow rate is modulated to l
maintain the radiation chemistry laboratory at a slightly negative pressure differential with I
respect to the adjacent access corridor. The exhaust air is unfiltered and directed to the plant i
vent by the auxiliary / annex building ventilation subsystem for monitoring of gaseous offsite I releases.
I I
ne temperature of the supply air is controlled by a temperature sensor located in the radiatior I
chemistry laboratory. When the radiation chemistry laboratory room air temperature is low.
I hot water valves to the supply air hot water heating coils are opened to maintain the room I
temperature within its normal design temperature range. He security rum and primary I
sample room temperature conditions will vary according to the demand for s ,pplemental heat I
in the radiation chemistry laboratory. A humidifier maintains the relative humidity in the I
areas served above 35 percent for personnel comfort during periods of low outside humidity I
conditions. De exhaust air from the radiation chemistry laboratory and primary sample room I
is continuously monitored by a smoke monitor located in the common exhaust air ductwork.
I The operating supply air handling unit is automatically shut down and the standby unit is I
started if the supply airflow rate is below a predetermined setpoint.
I Abnormal Plant Operation I
If high airbome radioactivity is detected in the exhaust air from the annex building (which I
includes the exhaust air from the areas served by the radiation chemistry laboratory ventilation I
subsystem), the annex building supply and exhaust air isolation dampers close and the I
radiation chemistry laboratory supply air handling unit fan is automatically shut down De I
containment air filtration system provides filtered exhaust to maintain the isolated zone at a I
slightly negative pressure differential with respect to the outside environment and adjacent I
unaffected plant areas. Other abnormal conditions causing closure of the annex building i
isolation dampers also shut down the radiation chemistry laboratory supply air handling unit !
I fans.
I I
If smoke is detected in the common exhaust air duct from the radiation chemistry laboratory I
and prunary sample room, an alarm is initiated in the trutin control room. He radiationl I
chemistry laboratory remains in operation unless plant operators detemune that there is a need I
to manually shut down the subsystem. In the event of a fire occurring within the areas served, :
1 the HVAC ductwork penetratmg fire barriers close if the local air temperature exceeds j i q predetermined serpoints. l I
Revision: 7 April 30,1996 g)L l3 9.4 34 ] Westingflouse I
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- TX CONFIRMATION REPORT ** AS OF APR 2 'S7 15:38 PAGE.01 APG00 DESIGN CERT DATE TIME T0/FROM MODE MIN /SEC PGS STATUS i 01 4/ 2 15:31 #23:NRC G3--S 07'26 13 OK l
i i
e.
l FAX to DINO SCALETTI April 2,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay Don Lindgren Gordon Israelson Bob Vijuk Brian McIntyre OPEN ITEM #213 (M9.1.2-1)
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 59 calendar days away (43 business days). The relevant documentation related to Open Item #213 (M9.1.2-1) is SSAR Subsection 9.1.1.2.1 (pertinent pages are attached). This material was submitted to you in Revision 7 of the SSAR on May 6,1996. 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 W:nters 412-374-5290 l
l IU
AP600 Open Item Tracking System Database: Executive Summ:ry Date: 4/2/97 Selection: btem noj hetween 213 And 213 Sorted by item 0 Item DSER Section/ Tule/Descripum Resp (W) NRC No. 11 ranch Questum Type Detail Status Engineer Status $tsus letter No / Dme 213 NRR/Sii.B 912 MTG-Oi Lindgten,DAsraelum Closed Actum W lM9.I.2-1 (SPENT FUEL STORAGE) The SSAR shoukt be updated to include the stmennt thz the spent fuel pod is setsnue Categwy I and is from internal nussdes _ j Ckved' SSAR. sectum 91.2.2. Rev. 3 mcludes the statement that the walls of'the spent fuelhud are an integral put of the senwnne Category I aunhary 'I building structure and thz the facility is protected from the effects of natural phenomena such as earthquakes, mind, tornados, ikxxh, a.ai external nussales.
I
'NRC - Actron N - check for sensnue classificaten
,Se am 9 I 2.2.1 descnbes the seisnuc deugn to prevent fuel rak failures which could prmhnce internal nusules, the evaluatum of dnyped fuel armt the ,
'scisnue quali ficaton of the fuel handlu g pb crane to prevent its failuie.
l l
!Sectum 9.l.1.2.1 meludes the statement thz the spent fuel area contains no credible sources of internal missiles.
Ckmed - The.e are no excepuuns to GDC 2 or 4 for the spent fuel pool The absence of nussile sources near the spent fuel pxn is discussed in the SSAR l
!and the ;xxd jtselfis a Sei_smic Cate6 cry I structure, _j M
-b u\
Page: 1 Total Records: 1
0
- 9. Auntiary Systems i
radiologically controlled area ventilation system, Section 11.5 for process radiation monitoring, subsection 9.1.3 for the spent fuel pool cooling system, and subsection 12.2.2 for airborne activity levels in the fuel handling area.
9.1.2.2.1 Spent Fuel Rack Design A. Design and Analysis of Spent Fuel Racks The spent fuel storage racks are purchased equipment. The purchase specification for t the spent fuel storage racks will require the vendor to perfonn conftrmatory dynamic and I stress analyses. The seismic and stress analyses of the spent fuel racks will consider the various conditions of full, partially filled, and empty fuel assembly loadings. The racks will be evaluated for the safe shutdown earthquake condition and seismic Category I requirements. A detailed stress analysis will be performed to verify the acceptability of the critical load components and paths under normal and faulted conditions. The racks I
rest on the pool floor and are evaluated to determine that under loading conditions they do not impact each other nor do they impact the pool walls.
The dynamic response of the fuel rack assembly during a seismic event is the condition which produces the governing loads and stresses on the structure.
Loads and Load Combinations The applied loads to the spent fuel racks are: '
=
Dead loads Live loads - effect of lifting the empty rack during installation
=
Seismic forces of the safe shutdown earthquake !
Fuel assembly drop analysis Fuel handling machine uplift - postulated stuck fuel assembly
=
Thermal loads i Table 9.1-1 shows loads and load combinations that are considered in the analyses of the )
spent fuel racks including those given in Reference 5.
)
The margins of safety for the racks in the multi-direction seismic event are produced using loads obtained from the seismic analysis based on the simultaneous application of three statistically independent, orthogonal accelerations.
B. Fuel Handling Machine Uplift Analysis An analysis will be performed to demonstrate that the racks can withstand a maximum uplift load of 5000 pounds. This load will be applied to a postulated stuck fuel assembly. Resultant rack stresses will be evaluated against the stress limits and will be demonstrated to be acceptable. It will also be demonstrated that there is no change in rack geometry of a magnitude which causes the criticality criterion to be violated.
Revision: 7 April 30,1996 p 5 9.1-8 T Westinghouse
t w
- 9. Auxiliary Systems C. Fuel Assembly Drop Accident Analysis In the unlikely event of dropping a fuel assembly, accidental deformation of the rack will be determired and evaluated in the criticality analysis to demonstrate that it does not cause the criticality criterion to be violated. The analysis will consider only the case of a dropped spent, irradiated fuel assembly in a flooded pool and will take credit for dissolved boron in the water.
For the analysis of a dropped fuel assembly,Iwo accident conditions are postulated. The first accident condition conservatively assumes that the weight of a fuel assembly, control rod assembly, and handling tool (2800 pounds total) impacts the top of the fuel rack from a drop height of 3 feet above the top of the rack. Both a straight drop and an inclined drop will be included in the assessment. Calculations will be performed which demonstrate that the impact energy is absorbed by the dropped fuel assembly, the rack cells, and the rack base plate assembly. Under these faulted conditions, credit is taken for dissolved boron in the pool water.
The second accident condition assumes that the dropped assembly and handling tool (2800 pounds) falls straight through an empty cell and impacts the rack base plate from a drop height of 3 feet above the top of the rack. The analysis will be performed which will demonstrate that the impact energy is absorbed by the fuel assembly and the rack base plate. At an interior rack location, base plate deformation is limited so that the pool liner is not impacted. At a support pad location, the stresses developed in the pool liner will be evaluated to be within allowable limits such that the liner integrity is maintained.
- Under these faulted conditions, credit is taken for dissolved boron in the pool water.
D. Fuel Rack Sliding and Overturning Analysis l
l Consistent with the criteria of Reference 5, the racks will be evaluated for overturning l and sliding displacement due to earthquake conditions under the various conditions of I full, partially filled, and empty fuel assembly loadings.
r E. Failure of the Fuel Handling Jib Crane The fuel handling jib crane is a seismic Category 11 component. De crane is evaluated to show that it does not collapse into the spent fuel pool as a result of a seismic event. :
i Stress analyses will be performed by the vendor using loads developed by the dynamic .
I analysis. Stresses will be calculated at critical sections of. the rack and compared ;
cceptance criteria referenced in ASME Section 111. Division l_ arm ME^^^ '
9.1.2.3 Safety Evaluation ne design and safety evaluation of the spent fuel racks is in accordance with Reference 5.
The racks, being Equipment Clus 3 and seismic Category I structures, are designed to W Westingh00S8 pf I Revision: 7 APril 30,1996 9.1 9
- 9. Auxiliary Spiems C. Fuel Assembly Drop Accident Analysis In the unlikely event of dropping a fuel assembly, accidental deformation of the rack will be determined and evaluated in the criticality analysis to demonstrate that it does not cause the criticality criterion to be siolated. The analysk considers only the case of a dropped new fuel assembly.
For the analysis of a dropped fuel assembly, two accident conditions are postulated. The first accident condition conservatisely assumes that the weight of a fuel assembly and handling tool (1625 pounds total) impacts the top of the fuel rack from a drep height of ;
3 feet. Both a straight drop and an inclined drop will be included in the assessment. !
Calculations will be performed which demonstrate that the impact energy is absorbed by !
the dropped fuel assembly, the rack cells, and the rack base ptrte assembly.
The second accident condition assumes that the dropped asserably and tool (1625 pounds) falls straight through an empty cell and impacts the rack base plate from a drop height of 3 feet above the top of the rack. An analysis will be performed that will demonstrate the impact energy is absorbed by the fuel assembly and the rack base plate. The resulting rack deformations will be evaluated in the criticality analysis to demonstrate that the criticality criteria are not violated.
D. Failure of the Fuel Handling Jib Crane l
t The fuel handling jib crane is a seismic Category 11 component. The crane and the attachment to the buildinc or"~"~ " '~-d !
'n anw that the crane does not fall into l J - 6 storage pit during a seismic event.
E. Internally Generated Missiles The fuel handling area does not contain any credible sources of intemally generated 1 I
missiles.
ess analyses will be performed by the vendor using loads developed by the dyna analysi l
~m will be calculated at critical sections of the rack and co to !
acceptance criteria referencea m nLC 5'6 III Division I Ash "%. i 9.1.1.3 Safety Evaluation i The rack, being a seismic Category 1 structure, is designed to withstand normal and postulated dead loads, live loads, loads resulting from thermal effects, and loads caused by the safe shutdown canhquake event.
The design of the rack is such that Ke rr remains less than or equal to 0.95 with new fuel of the maximum design basis enrichment. For a postulated accident condition of 11ooding of the new fuel storage area with unborated water, K err does not exceed 0.98.
Revision: 7 April 30,1996 9.1 4 (r) b- W W8511Dgh0US8 o
'o
+
- TX CONFIRMATION REPORT ** AS OF APR 2 '97 15:30 PAGE.01 APG00 DESIGN CERT DATE TIME T0/FROM NODE MIN /SEC PGS STATUS 01 4/ 2 15:27 301 504 2300 G3--S 03*03 05 OK l
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4 FAX to DINO SCALETTI .
I April 2,1997 I CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay Don Lindgren Gordon Israelson Bob Vijuk Brian McIntyre j l
OPEN ITEM #214 (M9.1.2-2) :
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 59 calendar days away (43 business days). The relevant documentation related to Open item #214 (M 9.1.2-2) is SSAR Subsection 9.1.2.2.1 (pertinent pages are attached). This material was submitted to you in Revision 7 of the' SSAR on May 6,1996. 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 i
V F
l g,
i
r r AP600 Open item Tracking System Database: Executive Summary Date: 4/2/97 Selection: [ item nol between 214 And 214 Sorted by item #
Item DSER Sectam/ Tule/I)esenpion Resp (W) NRC No Branch Questum Type Detal Staus Engineer Suzus Status letter No I Dse 214 NRR/SPLB 9. l .2 mig-OI Wong/BPC Ckmed Action W
~ ~ ^
[M9.l.2-2 (SPENT FUEL STORAGE) Provide a discumon on'the d'esign and a5M performance of cumponents locmed in tlA vic'inwy of the spent
- fuel storage pit, not designed to seisnuc Cawgory I standards and whose fahme could damage the fuel or safety 4elmed systerns and opipnent. The design jof these components shuuld enstue that they will not fail Annng a seisnuc event, are seisnucally restraned,or are temoved from the ami dwing normal
??: . . _ = .- z :_: -.2===~-=-~--:==== ~ = - . . .. : - ~~-z==
, Closed. SSAR, secuans 9.I.I 2 and 9. l.2.2, Rev. 3 include seasements that the spent fuel pool is an integral part of the seisnue Cacgory I auxihary hidang structure and that the facihty is protected from the effects of natural phenomena such as earthquakes, wind, tenudos, floods, and esternal numles.
The fuel handhng machine is a seisnue Casegory I component. These are no other components in the area of the spent fuel storage pool.
NRC- Acuan W / Actum W - see 204 jSectawn 912 2.1 descnbes the seisnuc quahfacation of the fuel handhng pb crane to prevent its future.
Closed - There are no excepuons to GDC 2 or 4 for the spent fuel pool. The almence of nussale sources near the spent fuel pool as discussed in the SSAR h the pool itselfis a Sensnuc Caegory I structure _ _ _ _ _ _ _ _ _ _ _ __ _ __
N L
S-i i
a Ibge: 1 Total Records: I
^
- 9. Auxiliary Systems radiologically controlled area ventilation system, Section 11.5 for process radiation monitoring, a
subsection 9.1.3 for the spent fuel pool cooling system. and subsection 12.2.2 for airborne activity levels in the fuel handling area..
9.1.2.2.1 Spent Fuel Rack Design A. Design and Analysis of Spent Fuel Racks He a fuel storage racks are purchased equipment. The purchase specification for bel storage racks will require the vendor to perform confirmatory dynamic and
- e. Ayses. The seismic and stress analyses of the spent fuel racks will consider the 4
varicus conditions of full, partially filled, and empty fuel assembly loadings. He racks will be evaluated for the safe shutdown earthquake condition and seismi:: Category I requirernents. A detailed stress analysis will be performed to verify the acceptability of the criticalload components and paths under normal and faulted conditions. He racks
! I rest on the pool floor and are evaluated to determine that under loading conditions they do not impact each other nor do they impact the pool walls.
The dynamic response of the fuel rack assembly during a seismic event is the condition which, produces the goveming loads and stresses on the structure.
Loads and Load Combinations The applied loads to the spent fuel racks are: O
- Dead loads
- Live loads - effect of lifting the empty rack during installation
=
Seismic forces of the safe shutdown earthquake
- Fuel assembly drop analysis Fuel handling machine uplift - postulated stuck fuel assembly
= Thermalloads Table 9.1-1 shows loads and load combinations that are considered in the analyses of the spent fuel racks including those given in Reference 5.
He margins of safety for the racks in the multi-direction seismic event are produced using los obtained from the seismic analysis based on the simultaneous application of three statistically independent, orthogonal accelerations.
B. Fuel Handling Machine Uplift Analysis An analysis will be performed to demonstrate that the racks can withstand a maximum j uplift load of 5000 pounds. His load will be applied to a postulated stuck fuel l
assembly. Resultant rack stresses will be evaluated against the stress limits and will be )
demonstrated to be acceptable. It will also be demonstrated that there is no change in rack geometry of a magnitude which causes the criticality criterion to be violated.
I Revision: 7 3 April 30,1996 9,1 8 %j Westinghouse i l
)
6
- 9. Auxiliary Systems C. Fuel Assembly Drop Accident Analysis In the unlikely event of dropping a fuel assembly, accidental deformation of the rack will be determined and evaluated in the criticality analysis to demonstrate that it does not cause the criticality criterion to be violated. The analysis will consider only the case of a dropped spent, irradiated fuel assembly in a flooded pool and will take credit for dissolved boron in the water.
For the analysis of a dropped fuel assembly, two accident conditions are postulated. The first accident condition conservatively assumes that the weight of a fuel assembly, control rod assembly, and handling tool (2800 pounds total) impacts the top of the fuel rack from a drop height of 3 feet above the top of the rack. Both a straight drop and an inclined drop will be included in the assessment. Calculations will be performed which demonstrate that the impact energy is absorbed by the dropped fuel assembly, the rack cells, and the rack base plate assembly. Under these faulted conditions, credit is taken for dissolved boron in the pool water.
The second accident condition assumes that the dropped assembly and handling tool (2800 pounds) falls straight through an empty cell and impacts the rack base plate from a drop height of 3 feet above the top of the rack. The analysis will be performed which will demonstrate that the impact energy is absorbed by the fuel assembly and the rack base plate. At an interior rack location, base plate deformation is limited so that the pool liner is not impacted. At a suppon pad location, the stresses developed in the pool liner will be evaluated to be within allowable limits such that the liner integrity is maintained.
Under these faulted conditions, credit is taken for dissolved boron in the pool water.
D. Fuel Rack Sliding and Overturning Analysis Consistent with the criteria of Reference 5, the racks will be evaluated for overturning and sliding displacement due to earthquake conditions under the various conditions of fg '; P. d amntv fuel assembly loadings.
E. Failure of the Fuel Handling Jib Crane ne fuel handling jib crane is a seismic Category 11 component. The crane is evaluate to show that it does not collapse into the spent fuel pool as a result of a seismic event.
Stress analyses will be performed by the vendor using loads developed by the dynamic analysis. Stresses will be calculated at critical sections of the rack and compared to acceptance criteria referenced in ASME Section III Division 1. Article NF3000.
9.1.2.3 Safety Evaluation he design and safety evaluation of the spent fuel racks is in accordance with Reference 5.
The racks, being Equipment Class 3 and seismic Category I structures, are designed to 8- Revision: 7 9.1-9 APril 30,1996
[ W85tingt100S6
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- TX CONFIRMATION REPORT ** AS OF APR 2 '97 15:26 PAGE.01 AP600 DESIGN CERT DATE TIME TO/FROM MODE MIN /SEC PGS STATUS 01 4/ 2 15:23 301 504 2300 G3--S 02"22 04 OK
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FAX to DINO SCALETTI I
April 2,1997 i
CC: Sharon or Dino, please make copies for: D. Jackson !
Ted Quay l Don Hutchings Bob Vijuk l Brian McIntyre i l
OPEN ITEM #3 ( RAI 410.261) i 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 59 calendar days away (43 business ;
days). In my quest to make sure we have provided NRC with everything needed to prepare an l FSER, I have been providing background packages for open items that we believe are complete. !
Relevant documentation related to Open Item #3 (RAI 410.261) is attached. Action was i completed on this Item by our submittal of letter NTD-NRC-97-4993 on February 21,1997 (over one month ago). A copy of this letter, with the pertinent attachments, is included with this fax. We believe the information in this letter completed our action on item #3 and request that NRC review the material we have provided and provide a definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed."
Thank you.
t:#^
Jim Winters 412-374-5290
Al%00 Open item Tracking System Database: Executive SummC87 Date: 4/2/97 Selection: [ item no] between 3 And 3 Sorted by item #
Item DSER Secton/ Title /Desenpron Resp (W) NRC No. Branch Question Type Detad Staus Enginees Stsus Staus tener No. / Dse 3 NRR/SPLB 9.2.10 RAIOf SCS/Hutchings Closed Aamm W NTD-NRC-94-4326 10/21/94
~~~ ~
~" ~ ~
[Qaestion 410 261 (fills.' Hot Water fleermgIystemi ' '
Prmule the following inforrnanon on the hot weer heating system (IlliS) discussed in Section 9210 of the SSAR:
- a. Specify the operanng pressures and L.m.h of the tills pping that supply hot wawr to major areas of the plant.
- b. Ase any of the HHS lines routed into the- = _.a that regiare _ - ~ 2 isolation? If so.specify !w safety class and setsnuc category of the portions of the Hil5 piping at the contanment penetsmions between the-- e ' 2 isolarxin valves.
le. Spufy the putential c---;-- -- es of a brealt in the HHS piping and the protective na:asures to prevent danage (such as ikunhng) on safety- reized systems. i
!d _ _Do.any of tne {IH_S hnes run over or through the control roomf _ _ _ _ _ _ __ , _ _ _ _ _ , __ _
~~ ' ~ ~
__ _ _ _ , _ j '~
[CI'dResponsc phavitied'via 5D- RUUI3263AR5v73'resisedid91IO1' jNRC Status Update provided in Septender 5.1996 letter
- In a lener dated lune 21.19% Westinghouse responded to RAl# 410 289. The staff reuewed this response and found n acceptable flowever,the staff
- requests that part of the response, such as items #1 and #2 to be included in the SSAR. Action Westinghouse Action W - Revision 10 of the SSAR included the requested informanon. Item #1 of RAI4410.289 is included in subseason 9 2.7 2.2. first paragraph of g " Component Desenpnon", for the Chilled Waer System. Item #2 is included as both a design basis (Section 9 2.1012. second bullet) and a design
- descnpaion (Section 9.2.10.2.1. fifth paragraph) for the Ika Water Heating System. RAI 410.289 naast be revised to reflect this SSAR mfarmatum.
k Action N - Response to RAI 410.289 was pros;ded as Revisioa I by NSD-NRC-97-4962 of I/30/97.
Action W - Revise response to RAI 410 261 to be consissent with response to RAI 410.289.
4 .Chwed - Response provided by NTD-NRC47-4993 of 2/21/97. _ _ __ ,_ _ __
Page: 1 Total Records: I
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4 ;
, -O
. Westinghouse Energy Systems Box ass Electric Corporation Pmswen Pensem '5230.cass NSD-NRC-97-4993 DPC/NRC0747 Docket No.: STN-52-003 February 21,1997 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 TO: T.R. QUAY
SUBJECT:
WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION ON THE AP600.
4
Dear Mr. Quay:
Enclosed a're three copies of the Westinghouse responses to open items on AP600 topics. Responses to nine RAls are included in this transmittal. RAI 410.261 provides infonnation on Section 9 of the SSAR. Responses to RAI 440.571, Revision 1, discusses the OSU Test Analysis Report. Responses
- to RAls 260.83,84,85,86,87,88, and 89 address questions on Section 3 of the SSAR.
r The NRC technical staff should 'eview these respenses as a part of their review of the AP600 design.
These responses close, from a Westinghouse perspective, the addressed questions. The NRC should .
inform Westinghouse of the status to be designated in the "NRC Status" column of the OITS.
Please contact Brian A. McIntyre on (412) 374-4334 if you have any questions concerning this transmittal.
A.
Brian A. McIntyre,- anager Advanced Plant Safety and Licensing .
jml Enclosures cc: T. Kenyon, NRC - (w/o enclosures)
W. Huffman, NRC - (w/ enclosures)
N. Liparuto, Westinghouse - (w/o enclosures)
A ._
l I
i l
l
. NRC REQUEST FOR ADDITIONAL INFORMATION l i
l Ouestion 410.261 Provide the following information on the hot water heating system (HHS) discussed in Section 9.2.10 of the SSAR:
- a. Specify the operating pressures and temperatures of the HHS piping that supply hot water to major areas of the plant.
- b. Are any of the HHS lines routed into the containment that require containment isolation? If so, specify the safety class and seismic category of the portions of the HHS piping at the containment penetrations between i the containment isolation valves.
i'
- c. Specify the potential consequences of a break in the HHS piping and the protective measures to prevent damage (such as flooding) on safety-related systems.
- d. Do any of the HHS lines run over or through the control room?
Response
4
- a. The operating pressure and temperature of the hot water heating system (VYS) is about 300T (supply),
. 2207 (return) and 120 psig. De piping design conditions are 3207 and 200 psig.
- b. No VYS lines are routed inside the containment.
I
- c. No VYS piping is routed in rooms that contain safety-related equipment. There are no adverse consequences on safety-related components or equipment due to postulated breaks in the VYS piping routed in nonsafety-related areas.
- d. The VYS lines are not routed over or through the main control room.
This response is consistent with Revision 10 of the SSAR and the Revision I response to RAI 410.289.
SSAR Revision:
None i l
410.261-1 3 Westingflouse g,y, j j l
M , l 1
l4
\ ,
I
' c,
- TX CONFIRMATION REPORT ** AS OF APR 2 '97 15:19 PAGE.01 AP600 DESIGN CERT DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 4/ 2 15:16 301 504 2300 G3--S 02"15 04 OK i
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y FAX to DINO SCALETTI April 2,1997 i
CC: Sharon or Dino, please make copies for: Diane Jackson .
Ted Quay Don Lindgren Gordon Israelson Bob Vijuk Brian McIntyre l OPEN ITEM #210 (M9.1.1-8)
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 59 calendar days away (43 business days). The relevant documentation related to Open Item #210 (M9.1.1-8) is SSAR Subsection 9.1.1.2 (pertinent pages are attached). This material was submitted to you in Revision 7 of the SSAR submitted to you on May 6,1996. 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
i l
4
/ 3 4
C- -
AP600 Open Itein Tracking System Database: Executiv2Summiary Date: 4/2/97 helecties: litem no] between 210 And 210 Sorted by item #
hem DSER Sation/ Resp Tule/thnpteun (W) ' NkC No. Branch Questumi -
Type Detal Status Engmeer Samus Staus ' s No-/ h 210 NRR/SPLB 91.1 ' MTG-Of Lindgren D. Closal Actum W
~
[M9$1.l-8 (NEW FUEL STOR GE) Ape' gell-Idthe SSAR$a referenceEmmle$o2bMi I8 ' 7 'Q^ Il l' 5);hmE~choide$e figwes j T AcadW7Eise'SSAR5becin55e[EE to'fiIgN.
~~ ~~ ' ' '
jNRC 3
Cloned - SSAR subsection 9.l.12. Revision 7, includes the proper references for relanwtship hetweca the new fuel storage faciley and other features of the Lfuel handhag area _ , , , , , _ _ _ _ , , , _ , , , , , _ _ _ _ _ _ , , , _ , _ _ _ _ , _ _ , _ _ _ __ __ , _ _ , _ , _
N M
u Page: 1 Total Records: 1
Fasass I
y 9. Auxiliary Syst:ms 9.1.1.2 Facilities Description l
De new fuel storage facility is located within the seismic Category I auxiliary building fuel handling area. He facility is protected from the effects of natural phenomena such as <
canhquakes, wind, tornados, floods, and external missiles by the extemal walls of the auxiliary I building. See Section 3.5 for additional discussion on protection from missiles. He facility ,
is designed to rnaintain its structural integrity following a safe shutdown earthquake and to i perform its intended function following a postulated event such as fire, internal missiles, or l pipe break. The walls surrounding the fuel handling area and new fuel storage pit protect the I fuel from missiles generated inside the auxiliary building. He fuel handling area does not contain a credible source of missiles. Refer to subsection 1.2.6 for a discussion of the auxiliary building. Refer to Section 3.8 for a discussion of the stmetural design of the new I fuel storage area. Refer to subsection 3.5.1 for a discu" inn nr miuile sources and protection. l e dry, unlined, approximately 15.5-feet deep reinforced concrete pit is designed to pros suppon for the new fuel storage rack. The rack is supported by the pit floor and laterally supported as required at the rack top grid structure by the pit wall structures. The walls of the new fuel pit are seismic Category I. He new fuel pit is normally covered to prevent foreign objects from entering the new fuel storage rack. Since the only crane that can access i the new fuel pit does have the capacity to lift heavy objects, as defined in subsection 9.1.5, the new fuel pit cover is not designed to protect the fuel assemblies from the effects of dropped heavy objects. Figures 1.2-7 through 1.2-10 show the relationship between the new fuel storage facility and other features of the fuel handling area.
l l ne new fuel storage pit is drained by gravity drains that are part of the radioactive waste I
drain system (subsection 9.3.5), draining to the waste holdup tanks which are part of the liquid I
radwaste system (Section 11.2). These drains preclode flooding of the pit by an accidental ase of water.
1 Nonseismic equipment in the vicinity of the new fuel storage racks is evaluated to confirm that its failure could not result in an increase of Kert beyond the maximum allowable Keff.
Refer to subsection 3.7.3.13 for a discussion of the nonseismic equipment evaluation.
A jib crane is used to load new fuel assemblies into the new fuel rack and transfer new fuel assemblies from the new fuel pit into the spent fuel pool. De capacity of the jib crane is limited to 2000 lbs. The new fuel pit is not accessed by the fuel handling machine or by the cask handling crane. This precludes the movement ofloads greater than fuel components over stored new fuel assemblies.
During fuel handling ope' rations, a ventilation system removes gaseous radioactivity from the atmosphere above the new fuel pit. Refer to subsection 9.4.3 for a discussion of the fuel handling area HVAC system and Section 11.5 for process radiation monitoring. Security for the new fuel assemblies is described in Section 13.6.
Revision: 8 June 19,1996 3 9.1 2 3 Westingt100S8 3
_. ___.- . . _ - - . . ._...._.-._.........._.._.._.___.__.m._..._..._
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- TX CONFIRMATION REPORT ** AS OF APR 2 '97 15:15 PAGE.01 AP600 DESIGN CERT l i
DATE TIME TO/FROM - MODE -MIN /SEC PGS STATUS 01 4/ 2 15:13 301 504 2300 G3--S 01'47 03 OK ,
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J. g FAX to DINO SCALETTI April 2,1997 CC: Sharon or Dino, please make copies for: Diane Jackson- r Ted Quay Don Lindgren Gordon Israelson Bob Vijuk Brian McIntyre OPEN ITEM #212 (M9.1.1-10)
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 59 calendar days away (43 business days). The relevant documentation related to Open Item #212 (M9.1.1-10) is SSAR Figures 1.2-8,1.2-9, and 9.1-1 enclosed herewith. This material was submitted to you in Revision 7 of the SSAR submitted :
to you on May 6,1996. 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."
l O
Jim Winters 412-374-5290 i
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AP600 Open Iten TrackingSysten Database: Executive Sumsnary - Date .#2/F7 helection: (item nol between 212 And 212 Sorted by item #
hem DSER Section/ Resp .NRC I Tule/ Description - : (W)
No_ Branch Questani Type Detal Status Engineer Status ' Status tener No, / Dese 212 NRR/SPl.B 9.1.1 MTG4)I . Landgeen.D Closed . Actant W
- M9.1 1 10(NEW ftEi 51TEh)[Provale Layoidi5awungs for the new f5hkva~uk asmifuel~~
~
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[Clewd - The vauk isA5e Rgure 1.2-9.' Fuel storage racks are vendur supphed isent A sketch of the fuel rack chis shown in 5gwe SSAR Figure .
- 9.1-1 Revision 5. No further revisions of the SSAR in these areas are anticipmeed +
I c
NRC - Action W - provede layout drawings - Rev. 8.
iClosed - SSAR Figmes 1.2-8 and 11-9. Revision 7, show the locanon of tihe new fuel storage area and the placement of the 56 new fuel racks. Fuct
~fsnorage -- -racks e are - a-vendor suppleed mem. A sketch of the fuel rack concept is shown i
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- TX CONFIRMATION REPORT ** AS OF APR 2 '97 15:12 PAGE.01 AP600 DESIGN CERT DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 4/ 2 15:10 301 504 2300 G3--S 02'26 05 OK 1
I 4
l
i FAX to DINO SCALETTI April 3,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay Don Lindgren Gordon Israelson Bob Vijuk
! Brian McIntyre .
OPEN ITEM #219 (M9.1.2-7)
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 58 cale'ndar days away (42 business days). The l relevant documentation related to Open Item #219 (M9.12-7) is SSAR Figures 1,2-7, 1.2-8, 1.2-9, l.2-10,1.2-11,1.2-13,1.214, 9.1-2, 9.1-3, and 9.1-4 enclosed herewith. This material was i submitted to you in Revision 7 of the SSAR submitted to you on May 6,1996. It is requested NRC l
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." ;
1 Jim Winters
! 412-374-5290 l
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Selecuen: [ item no] between 219 Arut 219 Sorted by item O -'
leem DSER Sectonf Resp ~NRC Tale /Descriguson (W)
No. Bram.h Queston Type Detad Status Engmeer Status Status 1.cner No. I Dae' 219 NRR/SPLB 9.1.2 MTG-Ol Imigren.D Cimed Actme W f
1 M9.I.2-7 (SPENT FUEL STI* AGE) i
^
,.D..A. I'YU"' Drawags for the spent fuel storage pet =. _-. _-_. - __
g Cimed The spent fuel storage pool and adpacent areas are shown on the general arangernent drawings. Refer to SSAR figure 11-9.- _ . . .
DISCUSSED AT I/25/95 MEETING BETWEEN WESTINGHOUSE AND NRC PLANT SYSTEMS BRANCH tNRC- Acton W - revise section 1.2 - Rev. 8.
Closed - SSAR Figures 1.2-7.1.2-8.1.2-9,1.2-10.1.2-11.1.2-11 and I .2-14. Revision 7, show the location of the spent fuel storage area and the placement :
of the 616 spent fuel racks. Fuel storage racks are a veador sigphed seem. A sketch of the fuel rack concept (rack and layout)is shown in the SSAR [
Figures 9.1-2,91-3, and 9 I-4. No further revisens of the SSAR in these aseas are anticipased.
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. 1. Introduction and G:neral Description of Plant 9"*
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,- h 1.2-43
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- 1. IEtroduction and General Description of Plant *manet
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l NuclearIsland General Arrangement i Plan at Elevation 135'-3 & 145' 9" l 1
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[ W85tillgh00S8 April 30,1996 f) p 1.2-45 1
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- 1. Introduction and General Description of Plant -. m m
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Plan at Elevation 153'-0" & 160'-6" I i
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- 1. Introduction rad Generai Description af Plant -
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- Section A-A ty Revision
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February 28,1997
- g3 b 1.2-51
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- 1. Introduedon and General Description of Plant -
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- I Revision: 5 I February 29,1996 T Westinghouse 9.1-53
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- 9. Auxiliary Systems l
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- 9. Auxiliary Systems a
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4
** TX CONFIRMATION REPORT ** AS OF APR 3 '97 16:11 PAGE.01 APG00 DESIGN CERT DATE TIME TO/FROM MODE MIH/SEC PGS STATUS 01 4/ 3 16:05 301 504 2300 G3--S 06'01 13 OK l
1 i i l I l l 1 i l l {
FAX to DINO SCALETTI April 3,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay - Don Lindgren Gordon Israelson Bob Vijuk Brian McIntyre OPEN ITEM #210 (M9.1.1-8) P 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 58 calendar days away (42 business days). The relevant documentation related to Open Item #210 (M9.1.1-8) is SSAR Subsection 9.1.1.2 (pertinent pages are attached). This material was submitted to you in Revision 7 of the SSAR submitted to you on May 6,1996 (over 11 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." f Jim Winters 412-374-5290 t I l i
. . - - _ . _ . . . . . . _ . . , - _ _ _ - - ~ _ _ - - --_--- - - - - --. - _ - - - - - - - -- - - - - - -- . ---.- - - - - - - - - - - - - - - - - - - - - - -
<. .~ .
AP600 Open Itens Tracking System Database: Executive Susumary page: v3/97 Sekction: [ item no] between 110 And 210 Sorted by item O leen - DSER Secuent Resp Tale /Descnpaon (W) NRC
& Branch Questen Type - Deta 1 Status Engineer Status Samus Wh I h*
210 NRR/SitB 9.11 MTG-OI Landgren.D. Okmed Acton W
~
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[M9.1.I 8 (NEW FUEL STORAGEion~page 9II-17 E'SSIR[a'refe'veSce is male so Figures 152 jil k.~l31-20 poEle copEcf EE fige E.} o- ++w> ema-an w msm an.-m m- >mmm m. m,mn_
,NRC- Actawi W - sevise SSAR to be ciasistent of reference so figures.
, Closed - SSAR subsecoon 9.1.1.2. Revision 7. includes the paper refesences for relanonship between the new fuel storage Scalay and other fessures of the fue! Ming area. _____ - _ __ _ _ _, .
5 r i I h I b !
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- 9. Auxill:ry Syst:ms
}a l 9.1.1.2 Facilities Description 4 The new fuel storage facility is located within the seismic Category I auxiliary building fuel handling area. The facility is protected from the effects of natural phenomena such as canhquakes, wind, tornados, floods, and extemal missiles by the extemal walls of the auxiliary building. See Section 3.5 for additional discussion on protection from missiles. The facility is designed to maintain its structural integrity following a safe shutdown earthquake and to perform its intended function following a postulated event such as fire, intemal missiles, or pipe break. The walls surrounding the fuel handling area and new fuel storage pit protect the fuel from missiles generated inside the auxiliary building. The fuel handling area does not contain a credible source of missiles. Refer to subsection 1.2.6 for a dia:ussion of the auxiliary boilding. Refer to Section 3.8 for a discussion of the structural design of the new fuel stene area. Refer to subsection 3.5.1 for a discussion of missile sources and protection. The dry, unlined, approximately 15.5-feet deep reinforced concrete pit is designed to provides suppon for the new fuel storage rack. He rack is supported by the pit floor and laterally supported as required at the rack top grid structure by the pit wall structures. The walls of , the new fuel pit are seismic Category I. - The new fuel pit is normally covered to prevent foreign objects from entering the new fuel storage rack. Since the only crane that can access . 1
-[
1 the new fuel pit does have the capacity to lift heavy objects, as defined in subsection 9.1.5, the new fuel pit cover is not designed to protect the fuel assemblies from the effects of y
/
dropped heavy objects. Figures 1.2-7 through 1.210 show the ielationship between the new fuel storage facility and other features of the fuel handling area. .- , l The ne bstorage pit.is-dramed by. gravity-drains that are part of the radioaclive waste I drain system (subsection 9.3.5), draining to the waste holdup tanks which are part of the liquid I radwaste system (Section 11.2). These drains preclude flooding of the pit by an accidental release of water. Nonseismic equipment in the vicinity of the new fuel storage racks is evaluated to confirm . that its failure could not result in an increase of Keftbeyond the maximum allowable Keft. Refer to subsection 3.7.3.13 for a discussion of the nonseismic equipment evaluation. A jib crane is used to load new fuel assemblies into the new fuel rack and transfer new fuel assemblies from the new fuel pit into the spent fuel pool. The capacity of the jib crane is limited to 2000 lbs. De new fuel pit is not accessed by the fuel handling machine or by the cask handling crane. His precludes the movement of loads greater than fuel components over stored new fuel assemblies. During fuel handling operations, a ventilation system removes gaseous radioactivity from the atmosphere above the new fuel pit. Refer to subsection 9.4.3 for a discussion of the fuel handling area HVAC system and Section 11.5 for process radiation monitoring. Security for the new fuel assemblies is described in Section 13.6. Revision: 8 June 19,1996 9.1 2 3 Westingt10USS Y5 . A
, r* ,.- 9. Auxilizry Systems j
9.1.1.2.1 New Fuel Rack Design A. Design and Analysis of the New Fuel Rack ' The new fuel storage racks are purchased equipment. The purchase specification for the new fuel storage racks will require the vendor to perform confirmatory dynamic and stress analyses. The seismic and stress analyses of the new fuel rack will consider the ' various conditions of full, partially filled, and empty fuel assembly loadings. The rack l-will be evaluated for the safe shutdown earthquake conditien against the seismic [ l Category I requirements. A stress analysis will be performed to verify the acceptability ' i of the critical load components and paths under normal and faulted conditions. The rack [ t rests on the pit floor and is braced as required to the pit wall stmetures. The dynamic response of the fuel rack assembly during a seismic event is the condition ( ' which produces the goveming loads and stresses on the structure. The new fuel storage ( rack is designed to meet the seismic Category I requirements of Regulatory Guide 1.29. 1 : Loads and I oad Combinations The applied loads to the new fuel rack are: - i e Dead loads ,
=
Live loads effect of lifting the empty rack during installation
=
Seismic forces of the safe shutdown canhquake
=
Fuel assembly drop accident Fuel handling jib crane uplift - postulated stuck fuel assembly ' Table 9.1-1 shows loads and load combinations considered in the analyses of the new fuel rack. ; The margins of safety for the rack in the multi-direction seismic event are produced using loads obtained from the seismic analysis based on the simultaneous application of three statistically independent, onhogonal accelerations. B. Fuel Handling Jib Crane Uplift Analysis An analysis will be performed to demonstrate that the racks can withstand a maximum uplift load of 2000 pounds. This load will be applied to a postulated stuck fuel assembly. Resultant rack stresses will be evaluated against the stress limits and will be demonstrated to be acceptable, it will also be demonstrated that there is no change in rack geometry of a magnitude which causes the criticality criterion to be violated. t Revision: 7 Y W86 thigh 00$8 9.l , APril 30,1996
'/ 5
o ..... _-
- 9. Auxiliary Systems I
C. Fuel Assembly Drop Accident Analysis in the unlikely event of dropping a fuel assembly, accidental deformation of the rack will be determined and evaluated in the criticality analysis to demonstiate that it does not cause the criticality criterion to be violated. The analysis considers only the case of a dropped new fuel assembly. For the analysis of a dropped fuel assembly, two accident conditions are postulated. The first accident condition conservatively assumes that the weight cf a fuel assembly and handling tool (1625 pounds total) impacts the top of the fuel rack from a drop height of 3 feet. Both a straight drop and an inclined drop will be included in the assessment. Calculations will be performed which demonstrate that the impact energy is absorbed by the dropped fuel assembly, the rack cells, and the rack base plate assembly. The second accident condition assumes that the dropped assembly and tool (1625 pounds) falls straight through an empty cell and impacts the rack base plate from a drop height of 3 feet above the top of the rack. An analysis will be performed that will demonstrate the impact energy is absorbed by the fuel assembly and the rack base plate. The resulting rack deformations will be evaluated in the criticality analysis to demonstrate that the criticality criteria are not violated. D. Failure of the Fuel Handling Jib Crane The fuel handling jib crane is a seismic Category 11 component. The crane and the attachment to the building structure is evaluated to show that the crane does not fall into the new fuel storage pit during a seismic event. E. Internally Generated Missiles The fuel handling area does not cortain any credible sources of intemally generated missiles. i Stress analyses will be performed by the vendor using loads developed by the dynamic i analysis. Stresses will be calculated at critical sections of the rack and compared to acceptance criteria referenced in ASME Section 111, Divisien I, Anicle NF3000.
)
9.1.1.3 Safety Evaluation The rack, being a seismic Category i structure, is designed to withstand normal and postulated dead loads, live loads, loads resulting from thermal effects, and loads caused by the safe shutdown eanhquake event. The design of the rack is such that Ke rr remains less than or equal to 0.95 with new fuel of the maximum design basis enrichment. For a postulated accident condition of floodmg of the new fuel storage area with unborated water, Kerr does not exceed 0.98. Revision: 7 April 30,1996 ){ 9.l.4 W Westinghouse (tr ==
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** TX CONFIRMATION REPORT ** AS OF APR 3 '97 16:04 PAGE.01 APG00 DESIGN CERT I
DATE TIME TO/FROM MODE MIN /SEC PGS STATUS 01 4/ 3 16:01 301 504 2300 G3--S 03'03 05 OK l 1 l l l l l I I l l l l l 1' l i l
! ) e I FAX to DINO SCALETTI . April 3,1997 l CC: Sharon or Dino, please make copies for: Diane Jackson ! Ted Quay J Don Lindgren Gordon Israelson Bob Vijuk Brian McIntyre OPEN ITEM #222 (M9.1.3-3) 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 59 caiendar days away (43 business , days). The relevant documentation related to Open Item #222 (M9.1.3-3) is SSAR Subsection 9.1.3.5 (pertinent pages are attached). This material was submitted to you in Revision 8 of the SSAR submitted to you in July 1996 (over 8 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." e Jim Winters 412-374-5290 t P L
AP600 Open Iteen T.acking Systesa Database: Executive Sameneary page: 4/3/97 Selecties: [ item no] between 222 And 222 Sorted by item # hem DSER Sectsonf Tale /Desenptum Resp (W) NRC No. Branch Questum Type- Detal Staeus Engmeer Samus Status W No. I Dee 222 NRR/SPLB v.13 MTGol Israelson Oosed Action W
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,M9 I 3-3 (SPENT 14]EL PIT COOLING SYS1TM)In accordance wah Table 9.1-4 of the SSAR the water level (m the spent fuel pn or what? Table -. . _ . .!
'shoukt be labeled) drops so 63 feet as a resuk of a seisnuc event inanediasely following a normal refueling. What is the nanunum height sequeed for
!slueldmg?, How is ,the minenmen height -
". and whm is the souru of adeup under these condnions? _ . _. ______ .
~
' CHESS R' Tide UI1and secnons 9.13.43 4 and 9.135. Rev. 3 desenbc the water levelin the spen 5t1 p[md after lu6s of cooling, station blackout f and seisnuc events. The sections define the mensmum waser level regiured for coohng the fuel and stase shat personnel are pruinhned Irusa canenng the - >
, spent fuel pet area when these as a loss of nonnel coobag. Radianon shieldmg normally provided by the waser above the fuel is em sequeed when normal I spes fuel pool coohng is not avalable. e
.NRC - Acuan W - revine 9.535 refesence to waser level Table 9.5-4. l l' .
@ . SSAR subsection 9.13 5. Revision 8. includes references to the SSAR Table 9.1-4.
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l , 9. Anrmary Systems o pool cooling must be terminated, sufficient time exists to allow for repairs of a leak in the system. 9.13.433 Loss of Offsite Power The spent fuel pool cooling system pumps are automatically loaded on the respective onsite standby diesel generator in the event of a loss of offsite power. The spent fuel pool cooling system is capable of providing spent fuel pool cooling following this event. 9.13.43.4 Station Blackout Following a loss of ac power (off-site power and both diesel generators), the heat capacity of the water in the pool is such that cooling of the fuel is maintained. Table 9.1-4 provides the l times before boiling would occur in the pool following station blackout for various scenarios as well as the muumum levels of water that would be reached. Activity releases due to pool boiling are analyzed. The release concentrations at the site boundary are small fractions of the limits specified in 10 CFR 20, Appendix B with no credit for removal of activity by building ventilation systems (which are not available during loss of ac power situations). See subsection 9.13.1. The equipment in the fuel handling area exposed to elevated temperature and humidity conditions as a result of pool boiling does not provide a safety-related mitigation of the effects of spent fuel pool boiling or station blackout. Oi I Spent fuel pool makeup for long term station blackout can be provided through a seismically qualified safety-related makeup connection from the passive contamment cooling system water storage tank. This connection is located in an area of the auxiliary building that can be accessed without exposing operating personnel to excessive levels of hadiation or adverse environmental conditions during boiling of the pool. Operating personnel are not required to enter the fuel handling area when normal cooling is not available, and are not required to enter the area to recover normal cooling. 9.13.5 Safety Evaluation ' i l
"Ihe only spent fuel pool cooling system safety-related functions are containment isolation and emergency makeup connections to the spent fuel pool. Containment isolation evaluation is described in subsection 6.23. The following provides the evaluation of the design of the
_ spent full pool as well as the ~' J _:.... e... ; ne spent fuel pool is designed such that a water level is maintained above the spent fue - assemblies for at least 72 hours following a loss of the spent fuel pool cooling system, and without makeup (see Table 9.1-4). The minimum water level to achieve sufficient I cooling is the sub-cooled, collapsed level (without vapor voids) required to cover the top I i of the fuel assemblies. i The heat load is assumed to be the heat load asrumed for a full con nff ' ' Revision: 11 W- higgse 3 9.1-19 February 28,1997
e a-3 I s'
- 9. Ammary Systems j The spen: fuel pool cooling system includes safety-related connections to establish makeup to the spent fuel pool within 72 hours following a design basis event including a seismic event.
Radiation shielding normally provided by the water above the fuel is not required when normal spent fuel pool cooling is not available. Personnel are not permitted in the area when the level in the pool is below the minimum level.
'Ihe acceptability of the design of the spent fuel pool cooling system is based on specific General Design Criteria (GDCs) and Regulatory Guides as described in Sections 3.1 and 1.9.
9.13.6 Inspection and Testing Requirements 1 Active components of the spent fuel pool cooling system are either m contmuous or ' intermittent use during normal system operation. Periodic visual inspection and preventive ', maintenance are conducted. No specific equipment tests are required since system components are normally in operation when spent fuel is stored in the fuel pool. Sampling of the fuel pool water for gross activity. l tritium and particulate matter is conducted periodically. 9.13.7 Instrumentation Requirements l
>1 The instrumentation provided for the spent fuel pool cooling system is discussed in the following paragraphs. Alarms and indications are provided as noted.
A. Temperature i Instrumentation is provided to measure the temperature of the water in the spent fuel pool and to give indication as well as annunciation in the main control room when nonnal temperatures are exceeded. Instrumentation is also provided to give indication of the temperature of the spent fuel
^
pool water as it leaves either heat exchanger. B. Pressure Instmmentation is provided to measure and give indication of the pressures in the spent fuel pool pump suction and discharge lines. Instrumentation is also provided at locations upstream and downstream from the spent fuel pool filter and demineralizer so that pressure differential across this equipment can be determined. High differential pressure across the spent fuel pool filter and deminemlizer is annunciated in the main control room. Revision: 11 February 28,1997 k 9.1-20 [ Westingh00S8
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** TX CONFIRMATION REPORT ** AS OF APR 3 '97 16:00 PAGE.01 ,
r APG00 DESIGN CERT L i DATE: ' TIME TO/FROM NODE MIN /SEC PGS STATUS ! 01 4/ 3 15:57 301 504.2300 G3--S 02"25 '04 OK i 1 9 P 9
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o l V l FAX to DINO SCALETTI April 3,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay ) Don Lindgren Gordon Israelson Bob Vijuk Brian McIntyre OPEN ITEM #216 (M9.1.2-4) 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 58 calendar days away (42 business days). The relevant documentation related to Open Item #216 (M9.1.2-4) is SSAR Subsection 9.1.4.2.2 (pertinent pages are attached). This material was submitted to you in Revision 8 of the SSAR in July 1996. 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 l 1 i Id'l
e ,sH AP600 Open item Tracking System Database: Executive Summary Date: N2/97 Selection: [ item nol between 216 And 216 S<wted by Item # hem DSER Sectionf Tithnption Re5P (W) P. No Branch Quesem Type Detail Status Eagmeet Status SFa W No I i)ase 216 NRR/5PLB 9,l .2 MTG-OI Israelw Closed n*
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!M9.12-4 (SPENT FUEL STORAGEi ~
;To demonstrade . - wah GDC 63 as relased to the monitanng of the status of the stosed spent f'uel. Wesunghunse should ascuss rahanon
- nwsueonag devices for prosecuan of personnel un the buelang contuusous ast inomtonng in the spent fuel area, and the avaslahnbry of umatermputte fCU"""_Qions bemn the fuel handhng rnadunes, refuebn$ machines and the control ror m _ _ _ _ _ _ , _ _ _ _ _ ,
'Cd I
I stumt comumcanons added to sutnecnon 9.l.422. Me 25 2E'se area nu~mstor5w the fuel 5and'Idarca and mEUnas'a ]
!punable imidre momear is used dunng fuel handing operanons.
'NRC. Acuan W - add poseer to I11 y - SSA_R sutnecuan 9 I 42.2. Revision 8. includes _a refema . :6AR acction Il.5 for _ratation momroring of re_fucimg opermyns._ .;
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^ ! l l radiation shield, as well as a reliable cooling medium for removal of decay heat. The boric acid concentration in the water is sufficient to preclude criticality. i The associated fuel handling stmetures may be generally divided into two areas: the refueling cavity which is flooded only during plant shutdown for refueling, and the sper.t fuel pool and transfer canal, which is kept full of water. See subsection 9.1.1.3 for new fuel assembly storage. The new and spent fuel storap areas are accessible to operating personnel. The refueling cavity and the fuel Storage arcs are connected by the fuel transfer tube which is fitted with a quick opening hatch on we canal end and a valve on the fuel storage area end. The hatch is in place except during refueling to provide containment integrity. Fuel is carried through the tube on an underwater transfer car. 1 Fuel is moved between the reactor vessel and the fuel transfer system by the refueling ! machine. 'Ibe fuel transfer system is used to move up to two fuel assemblies at a time
)
between the containment building and the auxiliary building fuel handling area. After a fuel l assembly is placed in the fuel container, the lifting arm pivots the fuel asst. ably to the , borizontal position for passage through the fuel transfer tube. After the transfer car transports { the fuel assembly through the transfer tube, the lifting arm at that end of the tube pivots the ) assembly to a vertical position so that the assembly can be lifted out of the fuel container. In the fuel handling area, fuel assemblies are moved about by the fuel handling machine. Initially, a short tool is used to handle new fuel assemblies, but the new fuel elevator must be used to lower the assembly to a depth at which the fuel handling machine can place the new fuel assemblies into or out of the spent fuel storage racks. Decay heat, generated by the spent fuel assemblies in the fuel pool is removed by the spent fuel pool cooling and cleanup system. After a sufficient decay period, the spent fuel assemblies are removed from the fuel tacks and loaded into a spent fuel shipping cask for removal from the site. 9.1.4.2.2 Refueling Procedure , New fuel assemblies received for refueling are removed one at a time from the shipping container and moved into the new fuel assembly inspection area. After inspection, the accepted new fuel assemblies are stored in the new fuel storage racks. For the initial core load, some new fuel assemblies may be stored in the spent fuel pool. Prior to initiating the refueling operation, the reactor coolant system (RCS) is boratad and cooled down to refueling shutdown conditions as specified in the Technical Specifications. Criticality protection for refueling operations is specified in the Technical Specifications. The following significant points are addressed by the refueling procedure: The refueling water and the reactor coolant contain approximately 2500 ppm boron. This concentration is sufficient to keep the core five percent Ak/k subcritical during the refueling operations.
] Revision: 11 W Westinghouse 9.1-23 February 28,1997
4 > l I
- 9. Auxillary Systems The water levd in LM .-ddia s-iy is high enougn to Keep the radiation I O!
a 155Te limits when the fuel assemblies are removed from the core. Radiation onitoring is described in Section 11.5.
=
Continuous commumcations am estaoiisned and maimamed between the main control room and the personnel engaged in fuel handling operations. One or more of the systems described in subsection 9.5.2 are used for this communication. The refueling operation is divided into four major phases: preparation. reactor disassembly, fuel handling, and reactor assembly. A general description of a typical refueling operation through these phases is provided below. { 9.1.4.2.2.1 Phase I - Preparation The reactor is shut down, borated, and cooled to refueling conditions (s 140*F) with a final l than 0.95 (all rods in). Following a radiation survey, the containment building is kert ess , enteird. At this time, the coolant level in the reactor vessel is lowered to a point slightly l below the vessel flange. The refueling machine console is removed from storage and placed on the refueling machine and cables are connected. Then, the fuel transfer equipment and ; refueling machine are checked for operation (subsection 9.1.4.4). 9.1.4.2.2.2 Phase II - Reactor Disassembly i Head cables are disconnected at the integrated head package (IHP) connector plate to allow removal of the vessel head. See subsection 3.9.7 for a discussion of the integrated head Olil package. The refueling cavity is prepared for flooding by checking the underwater lights, { tools, and fuel transfer system; closing the refueling cavity drain lines; and removing the hatch from the fuel transfer tube. With the refueling cavity prepared for flooding, the vessel head is unseated and raised above the vessel flange using the containment polar crane. See subsection 9.1.5 for requirements for the polar crane. Water from the in-containment refueling water storage tank (IRWST) is transferred into the refueling cavity by gravity and the spent fuel pool cooling system (See subsection 9.1.3). De vessel head and the water level in the refueling cavity are raised, keeping the water level just below the vessel head. When the water reaches a safe shielding depth (subsection 9.1.4.3.7), the vessel head is taken to its storage pedestal. The control rod drive shafts are disconnected. The internals lift rig is installed and the upper intemals are removed from the vessel. See subsection 9.1.5 for discussion of lifting rig requirements and design. The fuel assemblies are now free from obstructions, and the core is ready for refueling. 9.1.4.2.2.3 Phase III - Fuel Handling De refueling sequence is started with the refueling machine. The positions of partially spent assemblies are changed, and new assemblies are addeo to the core. The general fuel handling sequence is as follows: O Revision: 11 February 28,199; ' E% 9.1 24 3 Westinghouse
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** TX CONFIRMATION REPORT ** AS OF APR 3 '97 15:58 PAGE.01 AP600 DESIGN CERT i DATE TIME T0/FROM MODE T11N/SEC PGS STATUS l 01 4/ 3 15:54 301 504 2300 G3--S 02*42 04 OK l l
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9 FAX to DINO SCALETTI April 3,1997 CC: Sharon or Dino, please make copies for: Diane Jackson W. Huffman Don Lindgren Ted Quay Ed Cummins Bob Vijuk Brian McIntyre OPEN ITEM #153 (M5.2.5-9) THIS IS AN INFORMAL RESUBMITTAL OF INFORMATION TO OBTAIN NRC ACKNOWLEDGMENT OF RECEIPT. 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 58 calendar days away 42 business days. The relevant documentation related to Open Item #153 (M5.2.5-9) is attached to this fax. We provided our response for this item in my fax to W. Huffman on March 10, 1997. We believe that this information resolves the concerns of item #153. It seems a reasonable request that NRC acknowledge receipt of the information and that NRC has a responsibility to recognize that Westinghouse, as an applicant has submitted the requested information. We request that NRC provide a definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed." f 4 Jim Winters 412-374-5290 1 I i l
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AP600 Open Item Tracking System Database: Executive Summary Date: 4/3/97 Selection: [ item nol between 153 And 153 Sorted by item # Imn DSER Sectm/ Title /Descnytion Resp (W) NRC No. Branch Questson Type Detal Staus Engmeet Staus Samus letter No / Dse 153 NRR/SPLB 5.25 MTG.Ot Hutchmgs/BFC Ckned Adion W
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M* thefeasitivi'? =8 'eSPume_'=$ 8* equ'P'nent which nunusers _identifwd leakage. _ _ , _ _ _ _ _ _
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lames 5SAR liivI3 of SishmAE 5.2.5 idenufics the senAivddesectkm IEUm~ds7 ~ ~ ~ ~ ' ~_ _ _ , , _ _ l Actnon W - Sensativity was rennyved in subsequent SSAR Revisions. Provide a new response to this quesnon.
,Omed - SSAR nMup provided,to Huffinan by fam 3/W.
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f Page: 1 Total R *wwds: I -. ____ - __-. - _ - - _ - - . - . ~ . . ., ~~ . - - - . - - - . . . . . . . . - - - - - - - . . - .- - _ . - _ _ - - _ _ _ _ _ - _ - - _ _ _ - _ _ . _-
.' O i RECIPIENT INFORMATION SENDER INFORMATION DATE: /PIARc4 IO /997 NAME:
_ j,m Ww TO: LOCATION: StLL YuffrM^1 ENERGY CENTER - EAST i PHONE: FACSIMit.E: PHONE: Omee:///z.3 7y-s ago COMPANY: Facsimile: (i S pjdC win: 284 4887 outside: (412)374 4847 LOCATION: Cover + Pages 1+3 The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call: WIN: 284 5125 (Janice) or'Outside: (412)374 5125. COMMENTS: 0lL L A,mowo Ace iMmes or 3 ssM Aess wnut smato ne uve ar7s tw & I53 o^) IEC LMc DETrzinou sc-MserivoV. Tr usite aa evio SSAg
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- 5. Reactor Coolant Sptem and Connected Spiems nonnal level signify a possible increase in unidentined leakage rates and alert the plant operators that corrective action may be required. Similarly, increases in containment sump level signify an increase in unidentified leakage. The following sections outline the 1 methods used to collect and monitor unidentined leakage. ?
5.2.5.3.1 Containment Sump Level Monitor Leakage from the reactor coolant pressure boundary and other components not otherwise identined inside the contairunent will condense and flow by gravity via the floor drains and other drains to the contailunent sump. A leak in the primary system would result in reactor coolant flowing into the contaimnent sump. leakage is indicated by an increase in the sump level. The contaitunent sump level is monitored by two scismic Category I leves :,ensors. The level sensors are powered . from a safety related Class IE electrical source. These sensors remain functional when subjected to a safe shutdown earthquake in confonnance with the guidance in Regulatory Guide 1.45. The containment sump level and sump total flow sensors located on the discharge of the sump pump are part of the liquid radwaste system. Failure of one of the level sensors will still allow the calculation of a 0.5 gpm in-leakage rate within I hour, ne data display and processing system (DDS) computes the leakage rate and the plant control system (PLS) provides an alann in the main control mom if the average change in leak rate for any given measurement period exceeds 0.5 gpm for
, unidentified leakage.g Unidentified leakage is the total leakage minus the identified leakage. The leakage rate algorillun subtracts the identified leakage directed to the sump.
ne measurement interval must be long enough to pennit the measurement loop to adequately detect the increase in level that would correspond to 0.5 gpm leak rate, and yet ' short enough to ensure that such a leak rate is detected within an hour. The measurement interval is less than er equal to I hour. When the sump level increases to the high level setpoint, one of the sump pumps automatically starts to pump the accumulated liquid to the waste holdup tanks in the liquid radwaste system. The sump discharge flow is integrated and available for display in the control room. Procedures to identiff the leakage source upon a change in the unidentified leakage rate I into the sump include the following: Check for changes in containment a:nwsphere radiation raonitor indications. Check for changes in containment humidity, pressure, and temperature, Check makeup rate to the reactor coolant system for abnonnal increases,
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h m ,; o , , old t,dde luk )s o,o? GPH, L Revision: 10
,j J December 20,1996 ' 5.2-24 3 Westinghouse
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- 5. Reactor Coolant System and Connected Systems T' Ti '
iy a I fr.:ck for changes in water levels and other parameters in systems which could leak water into the contaisunent, and Review records for maintenance operations which may have discharged water into the containment. l 5.2.5.3.2 Reactur Coolant System Inventory Balance t Reactor coolant system inventory monitoring provides an indication of system leakage. Net level change in the pressurizer is indicative of system leakage. Monitoring net ' makeup from the chemical and volume control system and net collected leakage provides an imponant method of obtaining infonnation to establish a water inventory balance. An abnormal increase in makeup water requirements or a significant change in the water inventory balance can indicate increased system leakage. The reactor coolant system inventory balance is a quantitative inventory or mass balance calculation. This approach allows determination of both the type and magnitude of leakage. Steady-state operation is required to perfonn a proper inventory balance ' calculation. Steady-state is defined as stable reactor coolant system pressure, temperature, power level pmssurizer level, and reactor coolant drain tank and in-containment refueling water storage tank levels. The reactor coolant inventory balance is done on a periodic basis and when other indication and detection methods indicate a change in the leak rate. The mass balance involves isolating the reactor coolant system to the extent possible and observing the change in inventory which occurs over a known time period. This involves isolating the systems connected to the reactor coolant system. System inventory is determined by observing the level in the pressurizer. Compensation is provided for changes in plant conditions which affect water density. The change in the inventory determines the total reactor coolant system leak rate. Identified leakages are monitored (using the reactor coolant drain tank) to calculate a leakage rate and by munitoring the intersystem leakage. The unidentified leakage rate is then calculated by subtracting the identified leakage rate from the total reactor coolant system leakage rate. TA e, W,,;-
,ldt.Ct.& Le f t.d is 0,I 3 G P M -
Since the pressurizer inventory is controlled during normal plant operation through the level control system, the level in the pressurizer will be reasonably constant even if leakage exists. The mass contained in the pressurizer may fluctuate sufficiently, however, to have a significant effect on the calculated leak rate. The pressurizer mass calculation includes both the steam and water mass contributions. Changes in the reactor coolant system mass inventory are a result of changes in liquid density. Liquid density is a strong function of temperature and a lesser function of pressure. A mnge of temperatures exists throughout the reactor coolant system all of which may vary over time. A simplified, but acceptably accurate, model for determining mass changes is to assume all of the reactor coolant system is at T w Revision: 10 W Westinghouse E 5.2-25 December 20,1996 e b
- 5. Reactor coolant System and Connected Systems The inventory balance calculation is done by the data display and processing system with 1 additional input from sensors in the protection and safety monitoring system, chemical and volume control system, and liquid radwaste system. The use of components and sensors in systems required for plant operation provides confonnance with the regulatory guidance in Regulatory Guide 1.45 that leak detection should be provided following seismic events that do not require plant shutdown.
5.2.533 Containment Atmosphere Radioactivity Monitor Leakage from the reactor coolant pressure boundary will result in an increase in the radioactivity levels inside containment. The contailunent aunosphere is continuously monitored for airbome gaseous radioactivity. Air How through the monitor is provided by the suction created by a vacuum pump. Gaseous and Nn/F a concentration monitors indicate radiation concentrations in the containment atmosphere. s a.re , ud Q ar e_
.The_ gas-channel can-respot rapiGy o_reac4cr ccolaa! prerare Sc=da.ry Icakage. Nnha neutron activation product which proportional to power levels. AddhionallyrNyhas-a-
-relatively-short-half-life and consequently "i!! reackeqmtituiwn wpidly. An increase in activity inside contairunent would therefore indicate a leakage from the reactor coolant pressure boundary. Based on the concentration of Nn /Fa and the power level, reactor coolant pressure boundary leakage can be estimated.
w;ttacrth s e.0,5 GPH Ic.A H,3 /Q) ne No/Fo monitoring system ha: a !d;;h sc1Mty _wh_en_Jheacactor is operating at a power range higher than 20 percent. The monitor is seismic Category I. Conformance with the guidance that leak detection sho be provided following seismic events that do not require plant shutdown is provided by the seismic Category I classification. Safety-related Class IE power is not required since loss of power to the radiation monitor is not coruistent with continuing operation following an earthquake. fivdxa.a_m k m. m .m.., icr. :ca sicui U.S gpm T..m bc Ucmucd. Operating experience has n average long.tenn leakage (from sampling losses, collected leakoffs, and unidentified leakage to the containment) from the reactor coolant system ranges between 0.1 and 0.3 gpm. De No concentration will increase by at least 25 percent above an existing 0.1 gpm leakage background and almost 10 percent for an existing 03 gpm leakage. Both increases are well within the sensitivity of the Nn/Famonitor capabilities. I R:.dioactivity concentration indication and alarms for loss of sample flow, high radiation, and loss of indication are provided. Sample collection connections permit sample collection for laboratory analysis. The radiation monitor can be calibrated during power operation. 5.2.53.4 Containment Pressure, Temperature and Humidity Monitors Reactor coolant pressure boundary leakage increases containment pressure, temperature, and humidity, values available to the operator through the plant control system. r m/ pa p m % . mimiw m ddreL h Ie A is n & r a g: m J A co m ira.fi g sa & G 'y9.M2 ,
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** TX CONFIRMATION REPORT ** AS OF APR 3 'S7 15:51 PAGE.01 AP600 DESIGN CERT DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 4/ 3 15: 47 301 SO4 2300 G3--S 03*47 06 OK l
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s s t FAX to DINO SCALETTI l l April 3,1997 l CC: Sharon or Dino, please make copies for: Diane Jackson ) Ted Quay ' Don Lindgren j Gordon Israelson Bob Vijuk Brian McIntyre i OPEN ITEM #213 (M9.1.1-1) 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 58 calendar days away (42 business days). The relevant documentation related to Open Item #213 (M9.1.2-1) is SSAR Subsection 9.1.1.2.1 (pertinent pages are attached). This material was submitted to you in Revision 7 of the SSAR on May 6,1996. 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 1
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- AP600 Open item Tracking System Database: Executiv2 Summ;ry Date: 4/3/97 Selecties: [ item nol between 213 And 213 Sc:ted by item #
Item DSER Section/ Title /Desenption Resp (W) NRC No. Branch Questum Type Detal Staus Engineer - St.atus Status letter No. / Dase 213 NRR/SitB 91.2 MTG4M Lindgren.DAsraelson Ckwed Acnon W iM9.l .2-1 (SPENT FUEL STORAGE) The SSAR should be updmed to include the statement tha the spent fuel pool is seisnuc Casegory I and is protected
;from meernal nussales.
l : = . =:1 . : == === : := ===== = =- - = = =:- --. ==:.:= : :. _ : = 22:= = :- .- : - 22::.: : _= == :
; Closed. SSetR. sechon 9.1.2.2, Rev. 3 includes the htmement that the walls of the spent fuel pml are an integral pan of the seismic Category 6 aunihary
'inniding structure and that the facthry is prosected from the effects of natural phenomens such as carthquakes, wind, tomados, floods, and extemal nussdes.
jNRC- Acnon W - revise SSAR.
'NRC - Acnon N - check for sehnuc classification.
Secnon 91.2.2.8 descnbes the seisnue design to prevem fuel rack fadures ahich could produre intemal massdes, the evaluzion of dropped fuel and the seismic quehfication of the fuel handhng jib crane to prevent its failure. Secnon 9.1.1.2.1 includes the stmement that the sgent fuel area contains no credible sources of internal nussales.
;Chued - The e are no excepuons to GDC 2 or 4 for the spent fuel pool. The absence of nussale sources near the spent fuel pool is discussed in the SSAR '
(and the pel itself as a Seism Cmegory I samcasse , _ _ _ , _ _ , _ , _ , _ _ _ _ _ _ _ _ _ ,_ _ , _ , , , _ _ , _ _ _ , _ , _ N 9 v b
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- 9. Auxiliary Systems radiologically controlled area ventilation system, Section 11.5 for process radiation momtonng, P: i subsection 9.1.3 for the spent fuel pool cooling system, and subsection 12.2.2 for airborne '
activity levels in the fuel handling area. 9.1.2.2.1 Spent Fuel Rack Design A. Design and Analysis of Spent Fuel Racks
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The spent fuel storage racks are purchased equipment. The purchase specification for the spent fuel storage racks will require the vendor to perform confirmatory dynamic and stress analyses. 'llie seismic and stress analyses of the spent fuel racks will consider the j various conditions of full, partially filled, and empty fuel assembly loadings. The racks will be evaluated for the safe shutdown earthquake condition and seismic Category I i requirements. A detailed stress analysis will be performed to verify the acceptability of ' the critical load components and paths under normal and faulted conditions. The racks I I rest on the pool floor and are evaluated to determine that under loading conditions they do not impact each other nor do they impact the pool walls. The dynamic response of the fuel rack assembly during a seismic event is the condition which produces the governing loads and stresses on the structure. Loads and Load Combinations The applied loads to the spent fuel racks are: (-
=
Dead loads Live loads - effect of lifting the empty rack during installation
=
Seismic forces of the safe shutdown earthquake
=
Fuel assembly drop analysis
=
Fuel handling machine uplift - postulated stuck fuel assembly l
= Thermal loads Table 9.1-1 shows loads and load combinations that are considered in the analyses of the ;
spent fuel racks including those given ir. Reference 5. 1 The margms of safety for the racks in the multi-direction seismic event are produced using loads obtamed from the seismic analysis based on the simultaneous application of three statistically independent, orthogonal accelerations. B. Fuel Handling Machine Uplift Analysis I An analysis will be performed to demonstrate that the racks can withstand a maximum ; uplift load of 5000 pounds. His load will be applied to a postulated stuck fuel I assembly. Resultant rack stresses will be evaluated against the stress limits and will be demonstrated to be acceptable. It will also be demonstrated that there is no change in rack geornetry of a magnitude which causes the criticality criterion to be violated. Revision: 7 $ .,)3' April 30,1996 9,18 T Westinghouse
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- 9. Auxiliary Systems i
l C. Fuel Assembly Drop Accident Analysis in the unlikely event of dropping a fuel assembly, accidental deformation of the rack will be determined and evaluated in the criticality analysis to demonstrate that it does not cause the criticality criterion to be violated. The analysis will consider only the case of a dropped spent, irradiated fuel assembly in a flooded pool and will take credit for dissolved boron in the water. For the analysis of a dropped fuel assembly, two accident conditions are postulated. The first accident condition conservatively assumes that the weight of a fuel assembly, control ; rod assembly, and handling tool (2800 pounds total) impacts the top of the fuel rack from a drop height of 3 feet above the top of the rack. Both a straight drop and an inclined drop will be included in the assessment. Calculations will be performed which demonstrate that the impact energy is absorbed by the dropped fuel assembly, the rack cells, and the rack base plate assembly. Under these faulted conditions, credit is taken for dissolved boron in the pool water. The second accident condition assumes that the dropped assembly and handling tool (2800 pounds) falls straight through an empty cell and impacts the rack base plate from a drop height of 3 feet above the top of the rack. The analysis will be performed which will' demonstrate that the impact energy is absorbed by the fuel assembly and the rack i base plate. At an interior rack location, base plate deformation is limited so that the pool ' liner is not impacted. At a support pad location, the stresses developed in the pool liner vill be evaluated to be within allowable limits such that the liner integrity is maintained.
)
Und r these faulted conditions, credit is taken for dissolved boron in the pool water. 1 D. Fuel Rack Sliding and Overturning Analysis ! l Cor:sistent with the criteria of Reference 5, the racks will be evaluated for overturning ; and sliding displacement due to canhquake conditions under the various conditions of full, panially filled, and empty fuel assembly loadings. ' E. Failure of the Fuel Handling Jib Crane The fuel handling jib crane is a seismic Category 11 component. The crane is evaluated to show that it does not collapse into the spent fuel pool as a result of a seismic event. Stress analyses will be performed by the vendor using loads developed by the dynamic analysis. Stresses will be calculated at critical sections of the rack and compared to acceptance criterit, referenced in ASME Section III, Division I, Article NF3000. 9.1.2.3 Safety Evaluation The design and safety evaluation of the spent fuel racks is in accordance with Reference 5. The racks, being Equipment Class 3 and seismic Category I structures, are designed to
,i/
c /D Revision: 7 3 W85tkigh00S8 9.1-9 APril 30,1996
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- 9. Auxiliary Systems C.
F \ Fuel Assembly Drop Accident Analysis t in the unlikely event of dropping a fuel assembly, accidental deformation of the rack will I be determined and evaluated in the criticality analysis to demonstrate that it does not cause the criticality criterion to be siolated. The analysis considers only the case of a dropped new fuel assembly. I 1 For the analysis of a dropped fuel assembly, two accident conditions are postulated. The I first accident condition conservatively assumes that the weight of a fuel assembly and , handling tool (1625 pounds total) impacts the top of the fuel rack from a drop height of ' 3
~ feet. Both a straight drop and an inclined drop will be included in the assessment.
Calculations will be performed which demonstrate that the impact energy is absorbed by the dropped fuel assembly, the rack cells, and the rack base plate assembly. The second accident condition assumes that the dropped assembly and tool I (1625 pounds) falls straight through an empty cell and impacts the rack base plate from a drop height of 3 feet above the top of the rack. An analysis will be performed that will demonstrate the impact energy is absorbed by the fuel assembly and the rack base plate. The resulting rack deformations will be evaluated in the criticality analysis to i demonstrate that the criticality criteria are not violated. D. Failure of the Fuel Handling Jib Crane The fuel handling jib crane is a seismic Category 11 component. The crane and the attachment to the building structure is evaluated to show that the crane does not fall into the new fuel storage pit during a seismic event. E. Internally Generated Missiles The fuel handling area does not contain any credible sources of internally generr.ted missiles. Stress analyses will be performed by the vendor using loads developed by the dyna analysis. Stresses will be calculated at critical sections of the rack and com d to tance criteria referenced in ASME Section 111, Division 1. Artic! ' 9.1.1.3 Safety Evaluation The rack, being a seismic Category I structure, is designed to withstand normal and postulated dead loads, live loads, loads resulting from thermal effects, and loads caused by the safe
>hutdown earthquake event.
The design of the rack is such that Kerr remains less than or equal to 0.95 with new fuel of the maximum design basis enrichment. For a postulated accident condition of flooding of the new fuel storage area with unborated water, Kerr does not exceed 0.98. Revision: 7 April 30,1996
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** TX CONFIRMATION REPORT ** AS OF APR ~3 '97 15:46 PAGE.01 j t
AP600 DESIGN CERT
-DATE TINC TO/FROM MODE MIN /SEC PGS STATUS '
01 4/ 3 15: 43 301 504 2300 G3--S 02"30 04 OK I h i i f L I d 1
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s e FAX to DINO SCALETTI April 4,1997 CC: Sharon or Dino, please make copies for: D. Jackson Ted Quay Jeff Willis Don Hutchings Bob Vijuk Brian McIntyre OPEN ITEM #4 (RAI 410.262) , 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 57 calendar days away (41 business days). In my quest to make sure we have provided NRC with everything needed to prepare an FSER, I have been providing background packages for open items that we believe are complete. Relevant documentation related to Open Item #4 (RAI 410.262) is attached. Action was completed on this Item by our submittal of letter NTD-NRC-97-5022 on March 13,1997. A copy of this letter, with the pertinent attachments, is included with this fax. We believe the information in this letter completed our action on Item #4 and request that NRC review the material we have provided and provide a definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed." Thank you. Cbb Jim Winters 412-374-5290 i 1 W
c w AP600 Open Itens Tracking System Database: Executive Summary - Date: 4/4/97 Selection: [ item no] between 4 Aru.14 Sorted by Item O hem DSER Sectam/ . Tale /Descnpum Resp IW) NRC [ No. Branch - Questum Type Detad Status Engineer Status - Starus tener No. I Dane ! 4 NRR/SPLB 9.2.7 ' RAl-OI Waters Ckmal Actum W - NSD-NRC-97-5022 ! Question 410 262 (VWS, Central Clulled Waser System) , {The SSAR seases that the central clulled water system (VWS) is designed as a nonsetume system and is classified as Class D in Table 3.2-3 of the SSAR.
,However, the VWS is required to be funcnonal dunng full power and shutdown operatum to supply clulled watet to meet the coohng load demand of the plans HVAC systent To meet C-2 of RG 1.29,the system should be designed as scismic Casegory 11. Provide the reason why the VWS should be flassa_6ed as amenumc casegwA-._ _ __._ ,__-. _ _,_ ,_ .
. Resolved - Per note from J. Winners 08/19/96.
' Closed - In response to letter NSD-NRC-96 4817 dated Sept. 10,1996.
Action W - per emad from Jackson 12/16/96 Action W - Per telecon between Scaletti and Winters on 2/4/97, NRC needs a more defimtve response aboid the placement of non-seisnue ppe.
@ - See W-letter NSD-NRC-97_5022 dated 3/13/97. _ _ . _ .,_ _ _ __ _ _ _ _ _ _. _
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Westinghouse Energy Systems Electric Corporation sa ass . Pmsbutgn Pennsylvania 1$230 0355 NSD-NRC-97-5022 ! DPC/NRC0770 i Docket No.: STN-52-003 j
. March 13,1997 !
Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555
.i TO: T.R. QUAY l
SUBJECT:
WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL ; INFORMATION ON THE AP600. '
Dear Mr. Quay:
Enclosed are three copies of the Westinghouse responses to open items on AP600 topics. Responses. ! to five RAls are included in this transmittal. RAIs 410.262 Revision 1 and 471.23 Revision I respond - l to questions on the SSAR. RAI 440.567 responds to a question on the PRHR heat exchanger modeling. AP600 testing is addressed in RAI 440.578 and PRHR actuation is in 480.224. The NRC technical staff should review these responses as a part of their review of the AP600 design. These respenses close, from a Westinghouse perspective, the addressed questions. The NRC should , inform Westinghouse of the status to be designated in the "NRC Status" column of the OIT$. ! Please contact Brian A. McIntyre on (412) 374-4334 if you have any questions concerning this transmittal. sA-- f$ Brian A. McIntyre, Manager h i Advanced Plant Safety and Licensing i jml . 1 Enclosures l
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cc: T. Kenyon, NRC -(w/o Enclosures) W. Huffman, NRC - (w/ Enclosures) N. Liparulo, Westinghouse - (w/o Enclosures) 3ti?A i l
}
t* j l l l) . l NRC REQUEST FOR ADDITIONAL INFORMATION t.--- I Question 410.262 Revision 1
)
The SSAR states that the central chilled water system (VWS) is designed as a nonseismic system and is classified j as Class D in Table 3.2 3 of the SSAR. However, the VWS is required to be functional during full power and shutdown operation to supply chilled water to meet the cooling load demand of the plant 'HVAC system. To meet C-2 of RG 1.29, the system should be designed as seismic Category II. Provide the reason why the VWS should be classified as nonseismic category.
Response
The chilled water system is not required to support the function of safety related equipment. The main control room habitability control system includes passive cooling provisions for equipment required for safe shutdown. SSAR Section 9.2.7.3 indicates that portions of the system which are located in safety-related areas are designed such that a failure in the system will not unacceptably impact the operation of safety-related components. In general, the system is not located where failure could impact equipment required for safe shutdown. The obvious exception is in containment. From the containment penetration to the anchor at the peripheral support structure, chilled water system piping is designed as seismic Category II. Although not designed for seismic from the anchor on, the piping is supported by the peripheral support structure which is a seismic structure. This will prevent a failure of the chilled water system from adversely impacting the operation of safety-related SSCs. SSAR Revision: NONE gy 410.262-1 Rev. I kMN
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** TX CONFIRMATION REPORT ** AS OF APR 8 '97 14:02 PAGE.01 APSOO DESIGN CERT DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 4/ 8 13:59 301 504 2300 G3--S 03'15 04 OK
** TX CONFIRMATION REPORT ** AS OF APR 11 '97 10:17 PAGE.01 APG00 DESIGN CERT DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 4/11 10:16 #23:NRC G3--S 01'44 04 CK l
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FAX to DINO SCALETTI 5 April 4,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay Don Lindgren Gordon Israelson Richard Orr Bob Vijuk Brian McIntyre OPEN ITEM #207 (M9.1.1-5) 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 57 calendar days away (41 business days). The relevant documentation related to Open Item #207 (M9.1.1-5) is SSAR Subsection 9.1.1.2 (pertinent pages are attached). This material was submitted to you in Revision 7 of the
; SSAR on May 6,1996 (more than 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." ) Jim Winters ! 412-374-5290 N ,
C : AP600 Open Item Tracking System Database: Executive Summary , Date: 4/4/97 I Selection: [ item no] between 207 And 207 Sorted by item # hem DSER Sectionf Tale /Descnption Resp (W) NRC. , No. Brandi Quesuon Type retal Status E.7 Samus Status twer No. I Due 207 NRR/SPLB 9.1.1 MTG4)I Orr.R. Closed Action W
;M9.1.1-5 (NEW FUEL SlDRAGE)
Westanghouse should perform an analysis to ensure that the falure of non-setsnuc Category I systems or structures located in the viennety of the acw-fuel ' tsig mcy cannot cause an uusease m e,Q@g the maxnman abh W. _ _ _ _ _ _ _ _ , , _ _ - ~- _ __, , , _ , , _ _ _ _ _ , _
~ ~ ^
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Cb5715 eve are no'non-seisnuc stnactures or w..[A.u in the vicimty of the New Fuel pit DISCUSSED AT 1/25/95 MEETING BE1 WEEN WESTINGHOUSE AND NRC PLANT SYSTEMS BRANCit NRC - Actum W - Add possive stasement that no other sanscases / equspment to increase Keft
, Closed - SSAR subsection 91 12. Revision 7. includes a discussion of the evaluation of nonsessmic etynprnent in the vicinny of the new fuel storage r_acks.
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Ca e l f p 9. Auxillary Systems r I l 9.1.1.2 Facilitics Description ! I ne new fuel storage facility is located within the seismic Category I auxiliary building fuel l handling area. The facility is protected from the effects of natural phenomena such as I earthquakes, wind, tornados, floods, and external missiles by the extemal walls of the auxiliary building. See Section 3.5 for additional discussion on protection from missiles. The facility is designed to maintain its structural integrity following a safe shutdown earthquake and to perform its intended function following a postulated event such as fire, internal missiles, or pipe break. The walls surrounding the fuel handling area and new fuel storage pit protect the fuel from missiles generated inside the auxiliary building. De fuel handling area does not contain a credible source of missiles. Refer to subsection 1.2.6 for a discussion of the auxiliary building. Refer to Section 3.8 for a discussion of the structural design of the new fuel storage area. Refer to subsection 3.5.1 for a discussion of missile sources and protection. The dry, unlined, approximately 15.5-feet deep reinforced concrete pit is designed to provide support for the new fuel storage rack. The rack is supported by the pit floor and laterally supported as required at the rack top grid structure by the pit wall structures. He walls of the new fuel pit are seismic Category I. De new fuel pit is normally covered to prevent foreign objects from entering the new fuel storage rack. Since the only crane that can access I the new fuel pit does have the capacity to lift heavy objects, as defined in subsection 9.1.5, the new fuel pit cover is not designed to protect the fuel assemblies from the effects of dropped heavy objects. Figures 1.2-7 through 1.2-10 show the relationship between the new fuel storage facility and other features of the fuel handling area. I l The new fuel storage pit is drained by gravity drains that are part of the radioactive waste I drain system (subsection 9.3.5), draining to the waste holdup tanks which are part of the liquid I radwaste system (Section 11.2). These drains preclude flooding of the pit by an accidental release of water. __ onseismic equipment in the vicinity of the new fuel storage racks is evaluated to confirm i that its failure could not result in an increase of Kerr beyond the maximum allowable Kerr. Refer to subsection 3.7.3.13 for a discussion of the nonseismic equipment evaluation. A jib crane is used to load new fuel assemblies into the new fuel rack and transfer new fuel assemblies from the new fuel pit into the spent fuel pool. The capacity of the jib crane is limited to 2000 lbs. He new fuel pit is not accessed by the fuel handling machine or by the cask handling crane. His precludes the movement ofloads greater than fuel components over stored new fuel assemblies. Dunng fuel handling operations, a ventilation system removes gaseous radioactivity from the atmosphere above the new fuel pit. Refer to subsection 9.4.3 for a discussion of the fuel handling area HVAC system and Section 11.5 for process radiation monitoring. Security for the new fuel assemblies is described 'a Section 13.6. k_ Revision: 8 June 19,1996 9,1 2 [ W85tingh0US8
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** TX CONFIRMATION REPORT ** AS OF APR 8 '97 13:58 PAGE.01 '
i AP800 DESIGN CERT l l 1 DATE TIME T0/FROM MODE MIN /SEC PGS STATUS l 01 4/-'8 13:56 301 504 2300 G3--S 01 40 03 OK I i I l l l I
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h Westinghouse FAX COVER SHEET I D ? l RECIPIENT INFORMATION SENDER INFORMATION j l l DATE: 3/2,7/f7 NAME: 3, L uwgg j TO: LOCATION: ENERGY CENTER - 4 3 . [AeM o A/ EAST
- PHONE: FACSIMILE: PHONE: Office: 4, gy4 9g
! COMPANY: Facsimile: win: 284-4887 Mkd i outside: (412)374-4887 l LOCATION: Q ogg,ggg g i 4 l 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 call: l WIN: 284 5125 (Janice) or Outside: (412)374 5125. i COMMENTS: l T /A N.5 'T~Whi? P OL seaw /N 4 1.s A M AR k-up j OF T*H-E. S S A A. SS C77sw 3.1. %iS cussio At l O n> 7~NEAM kJ STA hT7 FI c A ha N. ~i]lI$ ist AA k- 4 0 ? l
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- 3. Design of Structures, Component *, Eq:1pment, and Systems l AP600 Conformance l
I As part of the Westinghouse Owners Group pror;*am on surge line thermal stratifica-l tion, Westinghouse collected surge line physical odgn and plant operational data for I all domestic Westinghouse PWRs. In addition, Westinghouse collected surge line I monitoring data from approximately 30 plants. This experience was used in the I development of the AP600 thermal stratification loadings. Monitoring of the AP600 I surge line is therefore not required. I I Request 2. d) l I Applicants are requested to update stress and fatigue analyses, as necessary, to ensure
! Code compliance. The analyses should be completed no later than one year after I issuance of the low power license.
I l AP600 Conformance l I Revision of the stress and fatigue analyses is not required for the AP600 surge line. I since the design analysis considers thermal stratincation and thermal striping. I l Request 3) l l Addressees are requested to generate records to document the development and I implementation of the program requested by Items 1 or 2, as well as any subsequent I corrective actions, and maintain these records in accordance with 10 CFR Part 50, 1 Appendix B and plant procedures. i l AP600 Conformance t i I AP600 procedures require documentation and maintenance of records in accordance I with 10 CFR Pan 50, Appendix B.
.ha!y:.;; cf the pre;:,uner nurg: !!n; ;; pdormed :c de.T n:.:: :: :h;; h app &d!:
requiremen:: cf :h: ASME See:ica !!! Ccd: r.m me: ";i ::aly;;; include ; censidera:ica uf phn: cp;m:ica, :hcrm;! :.::::i& :!cn and :hcrm;! ;:dping, u9 ng :r mpem:u= di::ribu::ca: and innsira::. " hid am developed frem :p;≠ on ::is:ing phn: meni: ?ng p c;mm:.. A monitoring program will be implemented by the Combined License holder at the first AP600 to record temperature distributions and thermal displacements of the surge line piping, as well as peninent plant parameters such as pressurizer temperature and level, hot a leg temperature, and reactor coolant pump status. Monitoring will be performed during hot functional testing and during the Grst fuel cycle. The resulting monitoring data will be evaluated to show that it is within the bounds of the analytical temperature distributions and displacements. o mi:nomn ii:2 032797 Revision: 12 W Westingt10US6 3.9-55 Draft,1997
0 Westliighouse FAX COVER SHEET , D , RECIPIENT INFORMATION SENDER INFORMATION DATE: 3,/27 /9 7 NAME: TO: LOCATION: ENERGY CENTER - D IA cAsced EAST PHONE: FACSIMILE: PHONE: Office:
- COMPANY: Facsimile: win: 284-4887 Aj A c., outside: (412)374-4887 LOCATION: 6 c>c q v u ,E_
Cover + Pages 1 + /y The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call: WIN: 284 5125 (Janice) or Outside: (412)374 5125. COMMENTS: D ANE , A77A &M Gb ARE 4 A AN6 FS Te 5 Ec77ad
- 3. I) A a2> A 7?B#b L-x 3h T~o kuvA Ess Ta2 cammsms 1m ouA v b id 7 &ALL F M LiFA T7Jis MEF/< . Iu 3L
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- 3. Design of Structures, Compone ts, Equipment cnd Systems 3.11 Environmental Qualification of Mechanical and Electrical Equipment I
This section presents information to demonstrate that the mechanical and electrical portions of the engineered safety features, the reactor protection systems, and selected portions of the : post-accident monitoring system are capable of performing their designated functions while exposed to applicable normal, abnormal, test, accident, and post-accident environmental I l conditions. The information presented includes identification of the equipment required to be environmentally qualified equip =a: and, for each item of equipment, the designated l functional requirements, definition of the applicable environmental parameters, and documentation of the qualification process employed to demonstrate the required environmental capability. The seismic qualification of mechanical and electrical equipment I is presented in Section 3.10. The ponions of post-accident monitoring equipment required to be environmentally qualified are identified in Table 7.5-1. 3.11.1 Equipment Identification and Environmental Conditions 3.11.1.1 Equipment Identification I A complete list of environmentally qualified electrical and mechanical equipment that is essential to emergency reactor shutdown, containment isolation, reactor core cooling, or containment and reactor heat removal, or that is otherwise essential in preventing significant release of radioactive material to the environment, is provided in Table 3.11-1. A list of environmentally qualified electrical and mechanical equipment and a summary of electrical and mechanical equipment qualification results are maintained as part of the equipment qualification file. The Combined License applicant is responsible for verification that the equipment qualification file is maintained during the equipment selection and procurement phase. 3.11.1.2 Definition of Environmental Conditions Appendix 3D identifies applicable normal, abnormal, and design basis accident environmental conditions conforming to General Design Criterion 4. These environmental conditions are associated with various plant areas by an environmental zone, as noted in Table 3D.5-1 and Table 3.11-1. For mild environments, the area conditions do not change as the result of an accident. There are no degrading environmental effects that lead to common mode failure of the equipment. The qualification of mechanical and electrical equipment located in a mild environment is demonstrated by conducting the plant surveillance activities carried out during the operational phase of the plant. The environmental conditions identified in Appendix 3D are defined as follows. Revision: 12 [ W85tifigh0US8 3.11-1 APril 30,1997
- 3. Desig; of Structures, Components, Equipment and Systems Normal operating environmental conditions are defined as those conditions existing during routine plant operations for which the equipment is expected to be available on a continuous basis to perform required functions.
Abnormal environmental conditions are those plant conditions for which the equipment is designed to operate for a period of time without accelerating normal periodic tests, inspections, and maintenance schedules for that equipment. The maximum and minimum conditions identified as the abnormal condition are based on the design limits for the affected areas. Design basis accident (DBA) and post-design basis accident conditions are those plant conditions resulting from various postulated equipment and piping failures during which the I i&ntined equipment identified in Table 3.11 must operate without impairment of the function. The design basis accident and post-design basis accident conditions are discussed in Appendix I 3D. Qualifica::c; is-bard en ec:npenen: :peciE =!cula:icn; th= necczay Compatibility of equipment with the specified environmental conditions is achieved by the following. Systems and components required to mitigate the consequences of a design basis accident or to perform safe shutdown operation are qualified to remain functional after exposure to the environmental conditions in Table 3D.5-5. Environmentally qualified equipment exposed to a harsh environment has a qualified life goal of 60 years. Demonstration of qualified life by test or test and analysis is provided by I equipn:=: ;uppli= :c the Combined License applicant, to address applicable aging effects. For critical components susceptible to aging, a qualified life is established that includes the effects of the total integrated radiation dose experienced at their respective locations within the plant. When a 60-year qualified life is not achievable, a shorter qualified life is established, and a replacement program is implemented. For equipment located in a mild environtnent, a design life goal is established by using known significant aging mechanisms and reliability data. Equipment qualification takes into account the most severe environmental conditions resulting from the design basis high-energy line break. Included in these conditions are the short-term peak transient temperature following a main steamline break (MSLB) and a radiation exposure , I and temperature due -to a loss of coolant accident (LOCA) within the reactor containment. Postulated high-energy line failures as defined in subsection 3.6.2.1.2 are assumed in areas where high-energy lines greater than 1 inch are routed. Essential equipment is protected j against the effects of jet impingement (subsection 3.6.2.4.1) and evaluated for spray effects if required (subsection 3.6.2.7). Revision: 12 April 30,1997 3.11 2 W W85tingh00S8
I e
- 3. Design of Structures, Compone:ts, Eq:1pme;t and Systems Active mechanical equipment is qualified for operability as discussed in subsection 3.9.3 and Section 3.10. This operability program, combined with the qualification of the electrical appurtenances (valve operators, solenoids, limit switches), demonstrates qualification under required environmental conditions. Active mechanical equipment is defined as equipment that performs a mechanical motion as pan of its safety-related function.
Nonactive mechanical equipment whose only safety function is structural integrity is designed according to ASME Code guidelines. The accident and post-accident environmental effects are considered in the design of such stmetural components as pump casings and valve bodies. The environmental qualification program is restricted to evaluating the design of critical nonmetallic subcomponents of active devices in a harsh environment, where failure results in loss of the active component. 3.11.1.3 Equipment Operability Times For the AP600 Class 1E electrical and active mechanical equipment, post-accident operability times are shown in Table 3D.4-2 in Appendix 3D. Specific information for each device qualified as part of the IEEE 323-1974 qualification program is contained in the appropriate equipment qualification data package. The active mechanical component is qualified for operability as discussed in Section 3.10, using test, analysis, or a combination of tests and analyses. This operability program, combined with the qualification of the electrical appurtenances (for example, valve operators) discussed in the appropriate equipment qualification data packages, demonstrates qualification. 3.11.1.4 Standard Review Plan Evaluation A discussion of the Standard Review Plan requirements in regard to environmental ! qualification of mechanical equipment is provided in subsection 1.9.2. 1 3.11.2 Qualification Tests and Analysis
- 3.11.2.1 Environmental Qualification of Electrical Equipment l
The AP600 approach for environmental qualification of Class IE equipment is outlined m Appendix 3D. This methodology is developed based on the guidelines provided in , IEEE 323-1974 (Reference 1), and 344-1987 (Reference 2). Qualification for equipment in a harsh environment is based on type testing or testing and analysis. Analysis may be used to determine significant aging mechanisms in mild ) environment applications. Type testing includes thermal and mechanical aging, radiation, and I exposure to extremes of environmental, seismic, and vibration effects. Type testing is done with representative samples of the production line equipment according to the sequence I Revision: 12 T Westinghouse 3.11 3 April 30,1997
r" O 3
- 3. Destga of Structures, Compone1ts, Equipment and Systems l
i indicated in IEEE 323-1974 to the specified service conditions, including margin. The testing l takes into account normal and abnorma! plant operation and design basis accident and post- l design basis accident operations, as required. i When reliable data and proven analytical methods are available, environmental qualification ; may be based on analysis supported by partial type test data. This method includes ^ justifi:ation of the methods, theories, and assumptions used (that is, mathematical or logical proof based on actual test data) that the equipment meets or exceeds its specified performance I requirements when subjected to normal, abnormal, and design basis accident environmental conditions. Regulatory guides providing guidance for meeting the requirements of 10CFR50, Appendix A, General Design Criteria 1, 4, 23, and 50; Appendix B, Criterion Ill to 10CFR50 and i 10CFR50.49, include Regulatory Guide 1.89, Regulatory Guide 1.30, Regulatory Guide 1.63, ) Regulatory Guide 1,73, Regulatory Guide 1.100, and Regulatory Guide 1.131. The ; maintenance surveillance program follows the guidance of Regulatory Guide 1.33. Additional information regarding conformance with each of these regulatory guides is given ! in Section 1.9. I 3.11.2.2 Environmental Qualification of Mechanical Equipment AP600 mechanical components identified in Table 3.11-1 are qualified by design to perform , their required functions under the appropriate environmental effects of normal, abnormal, I accident, and post-accident conditions as required by General Design Criterion 4 and discussed in Appendix 3D. For mild environments, the area conditions do not change as a result of an accident. There are no degrading environmental effects that lead to common mode failure of equipment in mild environments. Mechanical equipment located in harsh environmental zones ! is designed to perform under the appropriate environmental conditions. i For mechanical equipment, there are two categories of components: i l
=
Active equipment - equipment that performs a mechanical motion as part of its safety-related function. The program for environmental qualification of active mechanical components is based on a combination of design, test, and analysis of critical sub-components, which is supported by maintenance and surveillance programs.
- Nonactive equipment - equipment whose only safety-related function is structural integrity. Nonactive components are designed for structural integrity according to ASME Code, Section Ill, as discussed in Section 3.9.
Revision: 12 April 30,1997 3.11-4 [ WBStingtl00S8
I
- \
- 3. Design of Structures, Components, Eq:1pmert and Systems 3.11.3 Loss of Ventilation The abnormal environmental conditions shown on Tables 3D.5-3 and 3D.5-4 reflect anticipated maximum conditions based on loss of normal ventilation systems.
Normal containment heat removal is provided by the nonsafety-related containment air recirculation cooling system. If this system is out of service for an extended period of time, the passive containment cooling system may be initiated to maintain the temperature and pressure below the limits notel Environmentally qualified equipment located in containment performs its functions under these conditions until the normal containment cooling system is . restored. Equipment areas outside containment and outside the main control room are maintained at normal environmental conditions by nonsafety-related HVAC systems. If these systems are disabled. the heat generated by this equipment is absorbed by the surrounding concrete with an ambient temperature rise that does not exceed the abnormal condition. Normal HVAC is restored within 72 hours or temporary ventilation is provided as discussed in Section 6.4. If the normal nonsafety-related main control room HVAC is lost, the heat generated by equipment and people is absorbed by the surrounding concrete. Normal heating, ventilation, and air-conditioning is restored within 72 hours or temporary ventilation is provided as discussed in Section 6.4. 3.11.4 Estimated Radiation and Chemical Environment The plant-specific estimates of the radiation dose incurred by equipment during normal operation is shown in Table 3D.5-2 and the estimated doses following a loss-of-cociant accident are defined in Table 3D.5-5. j The identified equipment is qualified to perform functions in the radiation environments present during normal and design basis accident conditions. The normal operational exposure is based upon design source terms presented in Chapter 11 and subsection 12.2.1. The ; equipment and shielding configurations are presented in Section 12.3. Post-accident ' monitoring, reactor trip and engineered safety features system and component radiation l exposures are dependent on the location of the equipment in the plant. Source terms and other l accident parameters are presented in subsection 12.2.1 and Chapter 15.
'Ihe maximum combined integrated radiation dose inside containment is based on the effects l of the normally expected radiation environment (gamma) over the equipment's installed life plus that associated with the most severe design basis event (gamma and beta) during or following which the equipment is required to remain functional.
The chemical environment following a loss of coolant accident is primarily based on the chemistry of the reactor coolant system fluid since there is no caustic containment spray. Sump pH adjustments are considered for cenain qualification tests. This is discussed further in Appendix 3D. Revision: 12 3 W85filigh00S8 3.11-5 April 30,1997
- 3. Design of Structures, Components, Equipment and Systems 3.11.5 Combined License Information item for Equipment Qualincation File The Combined License applicant is responsible for the maintenance of the equipment qualincation file during the equipment selection and procurement phase.
3.11.6 References
- 1. IEEE 323-1974, "lEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations."
- 2. IEEE 344-1987, "IEEE Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations."
l Revision: 12 April 30,1997 3.11-6 [ Westirigt10US8
- l 3 l - ? i
- 3. Design of Structurr:. Components, Equipment and Systems l l
Table 3.11-1 (Sheet 2 of 44) 1 Environmentally Qualified Electrical and Mechanical Equipment Operating Envir. Time Qualification AP600 Zone Function Rrquired Program i Description Tag No. (Note 2) (Note 1) (Note 5) (Note 6) BATI'ERY CilARGERS IDSA Battery Charger IDSA DC l 2 ISOL 24 hr E IDSB Battery Charger IDSB DC 1 2 ISOL 24 hr E IDSB Battery Charger 2 IDSB DC 2 2 ISOL 72 hr E IDSC Battery Charger i IDSC DC 1 2 ISOL 24 hr E IDSC Battery Charger 2 IDSC DC 2 2 ISOL 72 hr E IDSD Battery Charger IDSD DC l 2 ISOL 24 hr E Spare Battery Charger IDSS DC I 2 ISOL 24 hr E DISTRIBUTION PANEIE IDSA L25 Vdc Dnt Panel IDSA DD 1 2 ESF 24 hr E IDSB 125 Vdc Dist Panel IDSB DD i 2 ESF 24 hr E IDSC 125 Vdc Dist Panel IDSC DD i 2 ESF 24 hr E IDSD 125 Vdc Dnt Panel IDSD DD 1 2 ESF 24 hr E IDSA 120 Vac Dist Panel i IDSA EA I 2 RT 5 min E i ESF 24 hr i PAMS 24 hr l IDSA 120 Vac Dist Panel 2 IDSA EA 2 2 . RT 5 nun E l ESF 24 hr l PAMS 24 hr l IDSB 120 Vac Dist Panel i IDSB EA I 2 RT 5 nun E I ESF 24 hr j PAMS 24 hr IDSB 120 Vac Dist Panel 2 IDSB EA 2 2 RT 5 nun E l l ESF 24 4Mir l l l PAMS I yr 43-tw IDSB 120 Vac Dist Paael 3 IDSB EA 3 2 PAMS I yr E i l l l i Revision: 12 April 30,1997 3.11-8 W Westinghouse
o
- 3. Design of Structrres, Compone:ts, Eq;ipment and Systems Table 3.11-1 (Sheet 3 of 44)
Environmentally Qualified Electrical and Mechanical Equipment Operating Ensir, Time Qualification AP600 Zone Function Required Program Description Tag No. (Note 2) (Note 1) (Note 5) (Note 6) IDSC 120 Vx Dist Panel i IDSC EA I 2 RT 5 nun E ESF 24 hr PAMS 24 hr IDSC 120 Vac Dist Panel 2 IDSC EA 2 2 RT 5 nun E ESF 24E Mr PAMS I yr M-he IDSC 120 Vac Dist Panel 3 IDSC EA 3 2 PAMS 1 yr E sDSD 120 Vac Dist Panel i IDSD EA 1 2 RT 5 nun E ESF 24 hr PAMS 24 hr IDSD 120 Vac Dist Panel 2 IDSD EA 2 2 RT 5 min E ESF 24 hr PAMS 24 hr FUSE PANELS IDSA Fuse Panel IDSA EA 4 2 ISOL 24 hr E IDSB Fuse Panel IDSB EA 4 2 ISOL 24 hr E IDSB Fuse Panel IDSB EA 5 2 ISOL 1 yr E l IDSB Fuse Panel IDSB EA 6 2 ISOL i yr 4-he E IDSC Fuse Panel IDSC EA 4 2 ISOL 24 hr E IDSC Fuse Panel IDSC EA 5 2 ISOL I yr E l IDSC Fuse Panel IDSC EA 6 2 ISOL I yr EMr E IDSD Fuse Panel IDSD EA 4 2 ISOL 24 hr E TRANSFER SMITCHES i l IDSA Fuced Transfer Switch Box 1 IDSA DF 1 2 RT $ nun E i ESF 24 hr PAMS 24 hr IDSB Fused Transfer Switch Box i IDSB DF I 2 RT 5 nun E ; ESF 24 hr l PAMS 24 hr IDSB Fused Transfer Switch Box 2 IDSB DF 2 2 RT 5 nun E ESF 24 hr PAMS 72 hr IDSC Fused Transfer Switch Box i IDSC DF i 2 RT 5 min E ESF 24 hr PAMS 24 hr IDSC Fused Transfer Switch Box 2 IDSC DF 2 2 RT 5 min E ESF 24 hr PAMS 72 hr IDSD Fused Transfer Swuch Box 1 IDSD DF 1 2 RT 5 nun E ESF 24 hr PAMS 24 hr IDSS Fused Transfer Switch Box 1 IDSS DF l 2 RT 5 nun E tSpare) ESF 24 hr PAMS 72 hr Revision: 12 3 Westingh00S8 3.11-9 April 30,1997
e
- 3. Design of Structures, Components, Eq:1pme:t and Systems Table 3.11-1 (Sheet 5 of 44)
Environmentally Qualified Electrical and Mechanical Equipment Operating Envir, Time Qualification AP600 Zone Function Required Program Description Tag No. (Note 2) (Note I) (Note 5) (Note 6) TRANSFOILMERS IDSA Regulating Transformer I IDSA DT I 2 ISOL 24 hr E IDSB Regulating Transformer i IDSB DT I 2 ISOL 72 hr E PAMS I yr IDSC Regulatmg Transformer i IDSC DT I 2 ISOL 72 hr E PAMS I yr IDSD Regulating Transformer i IDSD DT I 2 ISOL 24 hr E INVERTERS IDSA inverter IDSA DU i 2 RT 5 nun E ESF 24 hr PAMS 24 hr IDSB Inverter I IDSB DU l 2 RT $ min E ESF 24 hr PAMS 24 hr IDSB inverter 2 IDSB DU 2 2 RT 5 min E ESF 24 hr l PAMS I yr h IDSC Invener i IDSC DU I 2 RT 5 min E ESF 24 hr PAMS 24 hr IDSC Inverter 2 IDSC DU 2 2 RT $ min E ESF 24 hr l PAMS I yr h IDSD Inverter IDSD DU l 2 RT 5 nun E ESF 24 hr PAMS 24 ha SM1TCIIGEAR RCP 1 A 4160V Switchgear 51 ECS ES 51 2 ESF 5 nun E PAMS 2 wks RCP 1 A 4160V Switchgear 52 ECS ES 52 2 ESF 5 n3n E PAMS 2 wks RCP 2A 4160V Switchgear 53 ECS ES 53 2 ESF 5 nun E PAMS 2 wks RCP 2A 4160V Switchgear 54 ECS ES 54 2 ESF 5 nun E PAMS 2 wks RCP IB 4160V Switchgear 61 ECS ES 61 2 ESF 5 nun E PAMS 2 wks RCP (B dl60V Switchgear 62 ECS ES 62 2 ESF 5 nun E PAMS 2 wks RCP 2B 4160V Switchgear 63 ECS ES 63 2 ESF 5 nun E PAMS 2 wks RCP 2B 4160V Switchgear 64 ECS ES 64 2 ESF 5 nun E PAMS 2 wks Revision: 12 [ W8Stingt100S8 3.11-11 April 30,1997
- 3. Design of Structures, Components, Equipme:t, cnd Systems -
l l l l l APPENDIX 3D Methodology for Qualifying AP600 Safety Related Electrical and Mechanical Equipment l Safety-telated electrical equipment is tested under the environmental conditions expected to occur in the event of a design basis event. This testing provides a high degree of confidence I in the safety-related system performance under the limiting environmental conditions. Qualification criteria were revised by IEEE 323-1974 (Reference 1) and by Regulatory Guide 1.89, which endorses this IEEE standard. The concept of aging was highlighted in IEEE 323-1974, and interpretation of the scope of aging and implementation methods were subsequently developed.10CFR 50.49 provides the NRC requirements for qualification of equipment located in potentially harsh environments. Therefore, the guidance provided by IEEE 323-1974 is the evolutienary root of requirements, recommended methods, and quali6 cation procedures described in this appendix. Specific treatment of seismic qualification, pan of the qualification test sequence recommended in IEEE 323-1974, is addressed in IEEE 344-1987 (Reference 2). This appendix bases technical guidance, recommendations, and requirements for seismic qualification on IEEE 344-1987. The AP600 Equipment Qualification methodology addresses the expanded scope of IEEE 627-1980 (Reference 3), which encompasses the quali6 cation of Class 1E electrical and safety-related mechanical equipment. IEEE 627 generalizes the principles and technical guidance of IEEE 323 and 344. Compliance with the IEEE 323-1974 and 344-1987 is the specinc means of compliance with the intent of IEEE 627-1980 for safety-related electrical and mechanical equipment. Safety-related electrical and mechanical equipment is typically quali6ed using analysis, testing, or a combination of these methods. The specific method or methods used depend on the safety-related function of the equipment type to be qualified. Safety-related mechanical equipment, such as tanks and valves, is typically qualified by analysis, with supplementary l functional testing when functional operability is demonstrated only through testing, as is the ! I case for active valves. Either tTesting or testing combined with analysis is the preferre? l I method used for environmental and seismic quali6 cation of safety-related (Class 1E) electrical I equipment. The technical discussions of this appendix follow the format headings of the equipment qualification data packages (EQDPs) to be issued as specific qualification program ; documentation. This formatting (see Section 3D.7) permits easy cross-reference between the methodology defined in this report and the detailed plans contained in the equipment ' qualification data packages. Attachment A of this appendix is the format used for the equipment qualification data package. 1 Attachment B of this appendix, " Aging Evaluation Program," describes methods for addressing ' potential age-related, common-rnode failure mechanisms used in AP600 equipment a wnimoon ai2-03:797 Revision: 12 [ W8Sthigt10US0 3D-1 April 30,1997
- 3. Design of Structures, Components, Equipment, and Systems 3D.4.4 Test Sequence Where the test sequence deviates from that recommended by IEEE 323-1974, the deviation is justified. The test sequence employed for a given hardware item is specified in the equipment qualification data package Sections 2.1 and 3.6 (see Attachment A for example).
I Note that for this reference and subsequence references to Attachment A the information in l Attachment A will be completed by the Combined License applicant. Clarifications to the IEEE 323-1974 recommended test sequence are discussed in the following:
- 1. Burn-In Test For electronic equipment, a burn-in test is completed, before operational testing of the equipment, to eliminate infant failures. The test consists of energizing the equipment for a minimum of 50 hours at nominal voltage and frequency under ambient temperature conditions. Any malfunction observed during these tests are repaired, and the 50-hour burn-in test is repeated for the repaired portion of the equipment.
- 2. Performance Extremes Test For equipment where seismic testing has previously been completed employing the recommended methods ofIEEE 344-1987, seismic testing is not repeated. Testing of the equipment to demonstrate qualification at performance extremes is separately performed as permitted by IEEE 323-1974, subsection 6.3.2(3). Additional discussion is provided I in subsection 3D.6.5.1.
- 3. Aging Simulation and Testing j For equipment located in a mild environment, aging is addressed as described in subsections 3D.6.3,3D.6.4, and Attachment B. If there are no known aging mechanisms that significantly degrades the equipment during its service life, it is acceptable to perform seismic testing of unaged equipment. Separate testing or analysis (or both) is provided to demonstrate that the aging of components is not significant during the projected service or qualified life of the equipment.
- 4. Synergistic Effects An imponant consideration in the aging of equipment for harsh environment service is the possible existence of synergistic effects when multiple stress environments are applied simultaneously. This potential is addressed by conservatism inherent in the determination and use of the worst-case aging sequence and conservative accelerated aging parameters.
The combination of effects from pressure, temperatures, humidity, and chemistry are addressed by the high-energy line break (HELB) tests. Since the test item is not exposed to radiation during this test, the effects of this parameter are conservatively addressed by Revision: 12 owmmunm.onm April 30,1997 3 D-8 [ W85tingh0ljS8
- 3. Design of Structures, Components, Equipment, and Systems l
l f 3 D.4.8.3 Radiation An additional 10% is added to the calculated total integrated dose in specifying the test i requirements. l l 3D.4.8.4 Seismic Conditions Required response spectra included in Subsection 3.7.2 or other AP600 program specifications are the conditions to be enveloped. No amplitude margin is added to these conditions. Peak broadening is also discussed in Subsection 3.7.2. Seismic qualification by analysis addresses margin requirements by other methods of conservatism while using the same sets of requirements - no amplitude margin is included. For qualification tests, the test facility increases the amplitude of seismic profiles by 10 percent to incorporate margin. For most applications, considerable margin exists with respect to the acceleration levels employed and the width of the response spectra. Funher details are addressed in Attachment E. 3D.4.8.5 High Energy Line Break Conditions The envelopes specified for high-energy line breaks, in Figures 3D.5-2 and 3D.5-3, are selected to encompass the transients resulting from a spectra of loss of coolant accidents and high-energy line break sizes and locations, and various nodes in the containment. As a consequence, these design envelopes already contain significant margin with respect to any transient corresponding to a single break. The AP600 equipment qualification methodology requires that the qualification envelopes be derived with a margin of 15 F and 10 psi with respect to the design envelopes in i Figures 3D.5-2 and 3D.5-3 = :ha: hc i-ita:ini; : an=n: be repca:cd The margin on dose is identified by comparing the location specific dose requirements and the AP600 equipment qualification parameters. The alkalinity of the chemistry is increased by 10 percent with respect to the peak value determined for the AP600 containment sump conditions. 3D.4.9 Treatment of Failures The primary purpose of equipment qualification is to reduce the potential for common mode failures due to anticipated environmental and seismic conditions. The redundancy, diversity, and periodic testing of nuclear power plant safety-related equipment are designed to accommodate random failures of individual components. Where an adequate test sample is available, the failure of one component or device together with a successful test of two identical components or devices indicates a random failure mechanism, subject to an investigation concluding that the observed failure is not common mode. Where insufficient test samples prevent such a conclusion, any failures are investigated l l o kam12W30dn Rl2 032797 Revision: 12 W Westirighouse 3D-15 April 30,1997 1 l
l
- 3. Design of Structures, Compone1ts, Eq11pment, and Systems I
3D.4,10.1.3 Material Link This documentation certifies that the materials used in the equipment are represented in a materials aging analysis, such as that described in Attachment B, (Subprogram B). This link applies only to equipment whose equipment qualification data package references the materials aging analysis and reflects a comparison of the as-built drawings, baseline design document, or other documentation of the plant specific equipment to the materials aging analysis listing. 3D.4.10.2 Similarity Where differences exist between items of equipment, analysis may be employed to demonstrate that the test results obtained for one piece of equipment are applicable to a similar piece of equipment. Documentation of this analysis conforms with guidelines in IEEE 323 and 627, and Subsection 3D.6.2.1 and Section 3D.7 of this appendix. 3D.5 Design Specifications The conditions and parameters considered in the environmental and seismic qualification of AP600 safety-related equipment are separated into three categories: normal, abnormal, and design basis event. Normal conditions are those sets and ranges of plant conditions that are expected to occur regularly and for which plant equipment is expected to perform its safety-related function, as required, on a continuous, steady-state basis. Abnormal conditions refer to the extreme ranges of normal plant conditions for which the equipment is designed to operate for a period of time without any special calibration or maintenance effort. Design basis event conditions refers to environmental parameters to which the equipment may be I subjected without impairment of its defined operating characteristics for those conditions. Equipmen: mquired c cpernic hi!c ;ubjected :c :he design ba= cven: and .: cr.reme condi:icn; and if no: replaced, may aquim :ha: : :.: , inspec:icas, c.nd main:enance be pcFcrmed n :he equipmen:, befc= =:uming :c normal cpem::n; ecadi: em The following subsections define the basis for the normal, abnormal, design basis event, and post-design basis event environmental conditions specified for the qualification of safety-related equipment in the AP600 equipment qualification program. (these are cited in Section 1.7 of each equipment qualification data package; See Attachment A.) The service conditions simulated by the test plan are identified in equipment qualification data package Section 3.7. (See Subsection 3D.7.4.6 and .) In general, the parameters employed are selected to be equal to (normal and abnormal) or have margin (design basis event and post-design basis event) with respect to the specified service conditions of equipment l qualification data package, Section 1.7, as recommended by IEEE 323. These conditions are l conservatively derived to allow for possible alternative locations of equipment within the i plant. l l o Warrv12WOdn R12 012797 Revision: 12 [ W85tkigfl00S8 3D-17 April 30,1997
a ee y-.yN.- f~ -
-7 4 i i! 3. Desig 2 of Structures, Components, Eq;ipm:nt, ezd Systems Aenno Table 3D.4-3 (Sheet 2 of 2)
AP600 EQ PROGRAM MARGIN REQUIREMENTS Required Condition Parameter Margin Notes ACCIDENT: Transient T n:SA l Temperature 1) Temperature (+15'F) and pressure and Pressure (+10 psig peak) margins added to transient profile, e
- 2) F:;;&g : ;na:n: appEd : c;;; " -- n;;gir H l ::;np::=ur: r p ::,::ur:
Chemical effects +10% In alkalinity of adjusted sump pH. Not applicable outside containment. Radiation +10% Added to calculated total integrated dose. Submergence Note 1 Generally, precluded by design. Seismic / Il0% Of acceleration at equipment mounting Vibration point for either SSE or line-mounted equipment vibration. (See Subsection 3D.4.8.4.) Post-accident +10% In time demonstrated via Arrhenius Aging time / temperature relationship calculation. Note:
- 1. Margin in submergence conditions is achieved by increases in temperature (+15'F), pressure (+10%), and chemistry (+10% in alkalinity of adjusted sump pH). Also, accident conditions submergence testing envelops ;
abnormal conditions submergence conditions.
)
1 1 l l Revision: 12 o t-ni:nown men April 30,1997 3D-42 W W85tingh0USS
,e b Westinghouse FAX COVER SHEET D RECIPIENT INFORMATION SENDER INFORMATION DATE: 4-l '17 NAME: (4 4 c4 L4ntg TO: LOCATION: ENERdY CENN_R -
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EAST I PHONE: FACSIMILE: PHONE: Office: zha - p y_m 7 COMPANY: Facsimile: win: 284-4887 m #o R c_ outside: (412)374-4887 LOCATION: Cover + Pages 1+a .- 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: Tee - Mn c ked \s et no rs cken c e. 4d vit\ La ,ude -k & %9e n vn m ma ct cl % '1. % kcan. ~ n a v_e +l -ta (G . 44 -h
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l ,. I FAX to JOE SEBROSKY l 1 April 1,1997 ' JOE - PLEASE PASS A COPY TO: Bill Huffman
- j. Diane Jackson Dino Scaletti cc: Brian McIntyre ,
Ilm Winters i The NRC is requested to please acknowledge receipt of information related to each of the fallowing open items ; that are listed in the OITS. De list below includes those open items that Westinghouse believes is either :
" Closed", "Confrm-W", or "Confrm-N", ne NRC status column in OITS lists these items as " Action W." l Please advise Westinghouse of what should be placed in the NRC status column. If these are truely " Action W", I l please provide a description of the action Westinghouse is expected to take.
j To assist you, I've separated the item numbers by NRC Project Manager and topic. Bill Huffman:
- 1. Multiple SGTR: 1466,4525 '
- 2. T/H Uncertainty: 2960, 2961, 2981 - 2984, 2986, 3427 i Diane Jackson: 5 l 1. Seismic Margin Analysis: 2536 - 2542,'2792 - 2803, 3431 - 3436, 3485 - 3504, 5034 - 5039 l
l 2. PRA Chapter 42: 1482 (like 7%) 5 Joe Sebrosky: l i 1. Internal Fire Analysis: 1429, 1431, 3895 - 3897, 3944 - 3959, 4202 - 4212 i
- 2. Shutdown PRA: 1448,1450,2945,2958,2959,3009 l 3. Level 2/3 PRA & Severe Accident issues: 1652 - 1678, 1682 1689, 1691, 1692, 1695 - 1704, 17 % , 1707, 2141 2146, i
2148 - 2150, 2152 - 2154, 2156 - 2207, 2209 - 2219, 4123 - 4144. ' I l Thanks,
. l i
l Cyn la Haag l l 412-374-4277 i l l. i i I I l t l, -
FAX TO JOE SEBROSKY March 31,1997 cc: Barry Sloane Brian McIntyre Jim Winters l l Westinghouse status of three DSER Open items pertaining to the AP600 PRA internal fire analysis is being changed to " Closed" These are being statused as closed because the inf;,rmation requested by the DSER open item is covered by Chapte.r $7, Internal Fire Analysis, of Revision 8 of the AP600 PRA, and all follow-on questions and RAls related to the internal fire analysis have been answered by Westinghouse. The specific DSER Open items include: DSER 0119.1.3.2-5 (OITS #1429): Westinghouse should quantitatively -evaluate control room fires. Response - Control room fires are quantitatively evaluated in PRA Chapter 57 (Pav. 8). DSER 0119.1.3.2-6 (OITS #1430): Westinghouse should assess the risk of a fire induced loss of systems during shutdown conditions. Response - Shutdown conditions are evaluated in the internal fire analysis presented in PRA Chapter 57 (Rev. 8). DSER 01 19.1.3.2-7 (OITS #1431): Westinghouse should assess fire induced opening of the ADS valves in ths PRA. f Response - Fire-induced opening of the ADS valves is included in the internal fire analysis provided in PRA Chapter 57 (Rev. 8). The NRC Status column of the OITS currently has " Action W" for items #1429 and 1431. (OITS #1430 has
" Action N" is the NRC status colun.n). Please inform Westinghouse of what the NRC Status column should be for these open items.
Thanks, la Cynthia Haa 412 374-42 l t l l I 1 l l 1 1 1 l
AP. 00 Open Iteam Tracking Systems Databaw: Project M=agesnent Seminiary Datz: 4/2/97 Se:ection: [ item no] between 3I37 And 3137 Sorted by item 8 CoenFResp Engineer Title / Description
- hem DSER Sectson/ Issue Closure Path N (W) NRC ! l' No. Branch Questson Type Status Detail Res Est(hrs) Status gNCP DraA Review Transmit 3137 NRR/SRXB 21.6.l.7 3 DSEA4M Butler,L / arlin I Closed Action W f
21.6.1.7-5 \ LOFTRAN should not be applied to any analysis involving actuation of the ADS because it has not been benchrurked agemst ADS actuation experunents. Closed -The LOFTRAN code is not used to analyac any events involving ADS - - - The LOFTRAN code is used to evaluate the system reponse to an inadvertent opening of a pressurizer rehef vdve (e g, Pressurizer safety valse or ADS valve). The evaluation is limited to the initial portion of the event and is limited to the single > phaec steam flow portion only. O
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AP600 TECHNICAL SPECIFICATIONS s WESTINGHOUSE RESPONSES TO NRC QUESTIONS AND COMMENTS 15 I i
- 29) Draft TS 3.7.8, Secondary Coolant Leakage l The Plant Systems Branch has reviewed Westinghouse's letter, dated July 26, 1996, regarding "AP600 LBB QUESTIONS" and found the position described in the letter regarding steamline leakage control unacceptable. In the letter, Westinghouse revised its previous response to Q410.145 withdrawing its TS commitment for steamline leakage detection without withdrawing LBB for the l steamline application. The staff found that the alternative method proposed by Westinghouse, using administrative procedures, did not provide suilicient measures to justify LBB application for steamlines. By letter dated September 5, 1996, the staffindicated to Westinghouse that application of LBB to the steamlines is not acceptable without a TS for mainsteam leakage detection because the technology of LBB relies on the detection ofleaks prior to pipe breaks.
Westinghouse was requested to provide a proper TS for steamline leakage detection. In response, Westinghouse provided informal draft TS 3.7.8, Secondary Coolant
- Leakage on October 24,1996. The NRC staff (SPLB, TSB, and ECGB) art reviewing it. A telecon between Westinghouse and the stafTwas held on November 5,1996. The staff asked justification from Westinghouse as related to (1) the adequacy and margin of 5.0 gpm as the leakage limit and (2) the leakage reducing time of 8 hours before entering into Action B.1.
Response
(1) The leakage. limit has been conservative'v selected to correspond to a leakage crack size that is detectable, structurally stable even under seismic loads, and permits sufficient time for operator corrective action to preclude pipe rupture. The leakage limit (5.0 gpm)is 10 times the minimum leak dctection capability of the instrumentation. This large detection capability margin ensures that the leaks can be detected as assumed in the LBB analysis. The acceptability of the 5.0 gpm leakage limit is established by doubling the size of a 5.0 gpm leakage crack and verifying that the 2X crack is structurally stable and thus not liable to increase in size. Therefore, cracks twice as big as those allowed by the 5.0 gpm limit are stable and will not increase in size. (2) The Required Action A.18 hour Completion Time was selected based on the stability ofleakage cracks twice as long as those corresponding to the 5.0 gpm limit. If the leakage can be restored to within the 5.0 gpm limit within 8 hours, there is no technical basis for shutting the plant down to MODE 4 with RNS cooling, since the crack will not increase in size in any amount of time. If the leakage can not be restored to within the limit within 8 hours, it is considered ; that some difficulty has been encountered in attempting to control the leak and l that shutdown should be initiated so that repairs may be pursued.
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, i
j Main Steam Line 2-d ":in F 0dwater-14ne Leakage l B 3.7.8 i 3.7 PLANT SYSTEMS i 3.7.8 Main Steam Line and ":in Feedeat+e-Eine Leakage LCO 3.7.8 Main Steam Line and Maia F :dwater-tine leakage throu h the pipe walls inside containment shall be limited t . gp .
- S.O APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS , CONDITION REQUIRED ACTION COMPLETION TIME , A. Main Steam Line and A.1 Be in MODE 3. 8 hours f!aia reedeater Line leakage exceeds AND operational limit. A.2 Be in MODE 5. 48 hours i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1 Verify main steam line end meiw Per SR 3.4.8.1 f;;dwater-lin lea nto the containment sump OTS m.
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S.o l l 1 i l l (!)AP600 3.7-17 m:0348w.wof:1> t 10794 I i
/ - Main Steam Line And Nin Teedwater-Line. Leakage B 3.7.8 3.7 PLANT SYSTEMS 3.7.8 Main Steam Line.=d tir F00Mter Li= leakage BASES BACKGROUND A limit on leakage from the main steam " --*-
#^*+^r liner inside containment is required to limit system operation in the presence of excessive leakage. Leakage is limited to an amount which would not compromise safety consistent with the Leak-Before-Break (LBB) analysis discussed in Chapter 3 of the AP600 SSAR (Ref. 1). This leakage limit ensures appropriate action can be taken before the integrity of the lines is impaired.
LBB is an argument which allows elimination of design for dynamic load effects of postulated pipe breaks. The fundamental premise of LBB is that the materials used in nuclear plant piping are strong enough that even a large throughwall crack leaking well in excess of rates detectable by present leak detection systems would remain stable, and would not result in a double-ended guillotine break under maximum loading conditions. The benefit of LBB is the elimination of pipe whip restraints, jet impingement effects, subcompartment pressurization, and internal system blowdown loads. As described in Section 3.6 of the AP600 SSAR (Ref.1), LBB has been applied to the main steam line - " --" '- ""-- ' Here pipe runs inside containment. Hence, the potential safety significance of secondary side leaks inside containment requires detection and monitoring of leakage inside containment. This LCO protects the main steam lines and main feedwater lines inside containment against degradation, and helps assure that serious leaks will not develop. The consequences of violating this LC0 include the possibility of further degradation of the main steam lines, which may lead to pipe break. APPLICABLE The safety significance of plant leakage inside containment SAFETY ANALYSES varies depending on its source, rate, and duration. Therefore, detection and monitoring of plant leakage inside containment are necessary. This is accomplished via the instrumentation required by LCO 3.4.10. "RCS Leakage Detection Instrumentation," and the RCS water inventory balance (SR 3.4.8.1). Subtracting RCS leakage as well as (continued) b AP600 B 3.7-34 m A3349w.wpf 1>110700 d
Main Steam Line arid "ain feedweter L-ine Leakage B 3.7.8 i BASES APPLICABLE any other identified non-RCS leakage into the containment SAFETY ANALYSES area from the total plant leakage inside containment (continued) provides qualitative information to the operators regarding possible main steam line er ==4n feedwatee-+irie leakage. This allows the operators to take corrective action should leakage occur which is detrimental to the safety of the facility and/or the public. Although the main steam line and main f ter line leakage limit is not required by any-of-tgriteria 4 the NRC Policy Statementi this specification nas been included in Technical Specifications in accordance with NRC direction (Ref. 2). LC0 Main steam line or =ir, feedwater-1-ine leakage is defined as leakage inside containment in any portion of the two (2) 28" I.D. main steam line pipe walls or two (2) 16" I.D. main feedwater line pipe walls. Up toEO,$"gpm of leakage is allowable because it is within the capability of the makeup system and is wS4 below the leak rate for LBB analyzed cases of a main steam line er-ma4n=fe@=+ar 'ine crack twice as long as a crack leaking at ten (10) times the detectable leak rate under normal operating load conditions. Violation of this LC0 could result in continued degradation of the main steam or main feedwater lines. APPLICABILITY Because of elevated main steam system temperatures and pressures, the potential for main steam line er =in Jeemona ime leakage is greatest in MODES 1, 2, 3, and 4. In MODES 5 and 6, a main steam line nr ==ia ft:dwater lir4 leakage limit is not provided because the main steam system pressure is far lower, resulting in lower stresses and a reduced potential for leakage. In addition, the steam generators are not the primary method of RCS heat removal in MODES 5 and 6. (continued) h AP600 B 3.7-35 m:0340w vvpf 1>110794 e
Main Steam Line ed tir. rechetGr-L-ine Leakage J B 3.7.8 BASES (continued)
?
? q ACTIONS A.1 and A.2
)
i With main steam line o.t mam feeawater line leakage in i excess of the LC0 limit, the unit must be brought to lower l pressure conditions to reduce the severity of the leakage 1 and its potential consequences. Tne reactor must be placed. l in MODE 3 with 8 hours and MODE 5 within 48 hours. This 1 action reduces the main steam line pressure and leakage, and i ~ also reduces the factors which tend to degrade the main ^ steam lines. The Completion Time off hours f to reach MODE 3 i from full power without challenging plant systems is reasonable based on operating experience. Similarly, the Completion Time of 48 hours to reach MODE 5 without challenging plant systems is also reasonable based on 1 operating experience. In MODE 5, the pressure stresses acting on the main steam cr n ir, feedwater lines are much i lower, and further deterioration of the main steam or main : feedwater lines is less likely. SURVEILLANCE SR 3.7.8;l REQUIREMENTS Verifying that main steam line oe-man-feedwater-1-ine i leakage is within the LC0 limit assures the integrity of those lines inside containment is maintained. An early warning of main steam line.cr =ta fee 6::ter line leakage is l provided by the automatic system which monitor the { containment sump level. Main steam line er ::in feeinter ; The leakage would appear as unidentified leakage inside i containment via this system, and can only be positively identified by inspection. However, by performance of an RCS water inventory balance (SR 3.4.8.1) and evaluation of the cooling and chilled water systems inside.c tainment, determination of whether the main steam /' - in.feedwater Wes_aMotential source ( of unidentified leakage inside containment is possible. REFERENCES 1. AP600 SSAR, Section 3.6.
- 2. NRC letter, Diane T. Jackson to Westinghouse (Nicholas J.
Liparulo), dated September 5, 1996, " Staff Update to Draft Safety Evaluation Report (DSER) Open Items (01s) Regarding the Westinghouse AP600 Advanced Reactor Design," Open Item #365. h AP600 8 3.7-36 m M340w wpf1>110700
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I}4/0267 15816 to.039 DB2 ! a tee l f IJNITED STATES ! s* S NUCLEAR MEGULATORY COMMISSION wA&M6860 ton. o.C. asseMost h '.* ; { APPLICANT:' Westinghouse Electric Corporation ! FACILITY: AP600 a
- SUGJECT: SUptlARY Of AP600 MEETING REGARDING THE PASSIVE CONTAINMENT COOLING 1
i SYSTEM (pCS) A S MG0THIC COMPUTER CODE CLOSURE PATH FOR OPEN !
- ISSUES 4
j On March 5, Igg 7, the subject meeting was held at the U.S. Nuclear Regulatory Cennission (NRC) office in Rockville, Maryland. Robert Vijuk, Brian McIntyre, l and Jim Gresham of Westinghouse Electric Corporation (Westinghouse met with 4 Tim Martin, Gary Holahan, Carl Berlinger, Ted Quay, Diane Jackson,)and i 1 Tom Kenyon of the NRC staff. The purpose of the meeting was to discuss
! improved means for closure for the MGOTHIC computer code and PCS design i review.
1. li Mr. Quay opened the meeting stating that the progress for Westinghouse to complete its documentation and for the staff review has been slower than expected. The staff expressed a concern regarding an apparent lack of quality assurance review of the Westinghouse WG0THIC and PCS reports by Westinghouse prior to their submittal. Approximately 12 examples were discussed demonstrating the staff's concerns. F in house responded that the issue would be investigated. Additionally, k Tl house informed the staff that Mr. Gresham had been assigned as a full-time manager of the MGOTHIC and PCs review and Mr. Bruce Ra*1g was assigned as a licenr%g engineer. Westinghouse expressed a concern regarding the amount of questions on MG0THIC and th !:re! a /ecea Leer of NRC mana at review of the ques understood e concern, ;.;- .at?- Mk=The staff _ _ _ _ = ?r M stated that they and by management { that, to date, the staff's questions have all seemed reasonable. It as j agreed that increased communication, including additional meeti , would s assist the closure process. 1 A draft of this meeting sussary was provided stinghouse to allow them the i opportualty to ensure that the representa s of their consents and discussions were correct.
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Diane T. Jackson, reject Manager Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 52-003 i cc: See next page h/rJ $9:40 $NN O Oeme .2> dwar Car APR 2 '97 14: 10 PAGE.002 ( s.__ - d
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~%RRWW s u i N T E R \ APR - 31997 ;
$rian A. Mc Inty i i O F Fi C E a
To: B. McIntyre (NRC Informal Correspondence), R. Osterrieder, E. Carlin, File 7.6 l From: Earl H. Novendstern i
Subject:
LOOP and NOTRUMP l Date: March 31,1997 i i 1 ! Brian, ! 1 l
- Attached are draft copies of information given to Bill Huffman at our 3/28/97 ACRS meeting. l I ,
1 i j l I 1 1 l e i i d
)
I 1 I I l l i i 1 i 4 4 i From the desk of.. Ear 1H. Novendetem 1 Manager. Advanced and VVER
- Plant Safety Analysss Westnghouse Electne Corporation Box 355 Pittsburgh, PA 15230-0355 Phone: (412) 374 4790 )
Fax: (412) 374-4011 l 1 a
1 1 Dpu, Guo 7 u. g ,,m a .s/w/p7 I NOTRUMP RELATED ITEMS FOR WHICH WESTINGHOUSE BELIEVES THE (W) STATUS l SHOULD BE CHANGED TO CLOSED IN OITS l (3/27/97 Page 1) l ITEM 3229 DSER CN 21.6.2.4 3 1 Westinghouse needs to verify that the NOTRUMP code does not use the Bjornard and Griffith modification. Response: The NOTRUMP code does not use the Bjornard and Griffith modification to the i Zuber critical heat flux correlation. ITEM 3232 DSER CN 21.6.2.7-1 Comparisons of the NOTRUMP code simulations to the OSU and SPES-2 test data in the FV . report should confirm the applicability or insensitivity of the NOTRUMP low regimes models to the key system response parameters. > Response: The comparisons of the NOTRUMP code simulations to the OSU and SPES-2 test data have been included in the NOTRUMP Final Validation Report, WCAP 14807, Revision 1, which has been submitted via Westinghouse letter NSD-NRC-97-4960, dated 1/31/97. 1 J l I I l I
i Y NOTRUMP RELATED ITEMS FOR WHICH WESTINGHOUSE BELIEVES THE (W) STATUS SHOULD BE CHANGED TO CLOSED IN OITS (3/27/97 Page 2) ITEM 3140 DSEROI 21.6.2.2 1 l Westinghouse needs to identify which information from the NOTRUMP related RAI response will be formally incorporated into NOTRUMP related documentation such as the final verification and validation report, the code applicability document (WCAP-14206), or the SSAR. Response: Westinghouse has agreed to include all of the RAls and responses in the NOTRUMP Final Validation Report as an appendix. In addition, a table will be included in the ! report that briefly summarizes each RAI and indicates where in the report the response is l
, included (the RAI appendix and any other section of the report if it is included there also). i ITEM 3141 DSER-Ol 21.6.2.2 2 Westinghouse needs to submit the final verification and v::lidation report.
Response: The NOTRUMP Final Validation Report, WCAP-14807, Revision 1, has been submitted via Westinghouse letter NSD NRC-97 4960, dated 1/31/97. ; ITEM 3142 DSER-Ol 21.6.2.4-1 Westinghouse needs to explain what provision will be used to ensure that volumetric based , momentum equations will be used for all AP600 calculations. l t Response: The analysts responsible for developing the input for the AP600 plant calculations were on the team of people who performed the analyses contained in the NOTRUMP Final r Validation Report. As we made any changes to model related inputs such as the use of l volumetric based momentum equations, these changes were included in the AP600 plant j model to be used for SAR calculations as well as the calculation notes documenting the ' AP600 plant input. The AP600 input deck calculation notes are also independently reviewed I to make sure the appropriate inputs are used. j ITEM 3143 DSER-Ol 21.6.2.4-2 Westinghouse needs to submit the NOTRUMP assessment cases to demonstrate the adequacy of the re-casting of the momentum equation and drift flux equations in not volumetric form. Response: The assessment cases are included in Section 3.5 of the NOTRUMP Final Validation Report, WCAP-14807, Revision 1, which has been submitted via Westinghouse letter NSD NRC 97 4960, dated 1/31/97. l
i NOTRUMP RELATED ITEMS FOR WHICH WESTINGHOUSE BELIEVES THE (W) STATUS SHOULD BE CHANGED TO CLOSED IN OITS (3/27/97 Page 3) ITEM 3145 DSER-Ol 21.6.2.4-4 Westinghouse needs to explain what provision will be used in NOTRUMP to ensure that options to override the default flow partitioning will be used for all AP600 calculations. Response: ' The analysts responsible for developing the input deck (including choice of input options) for the AP600 plant calculations were on the team of people who performed the analyses contained in the NOTRUMP Final Validation Report. As we made any changes to model-related inputs, these changes were included in the AP600 plant model to be used for SAR calculations as well as the calculation notes documenting the AP600 plant input. The AP600 input deck calculation notes are also independently reviewed to make sure the appropriate inputs are used. ITEM 3146 DSER-Ol 21.6.2.4-5 Westinghouse needs to complete all benchmark and assessment calculations (to be included in the FV&V report) to demonstrate the acceptability of the logic modifications for application of the NOTRUMP code to the AP600 SBLOCA. Response: The assessment cases are included in the NOTRUMP Final Validation Report, WCAP-14807. Revision 1, which has been submitted via Westinghouse letter NSD-NRC 97-4960, dated 1/31/97. ITEM 3151 DSER-Ol 21.6.2.4 10 Westinghouse needs to submit benchmark calculations to demonstrate the acceptability of the adequacy of the NOTRUMP birthing logic, and its applicability to the AP600 SBLOCA. Response: After the preliminary calculations, this model was no longer used. The preliminary calculations were redone without the use of this model for inclusion in the NOTRUMP Final Validation Report. As a result, no benchmark was performed. ITEM 3153 DSER-OI 21.6.2.4-12 The NOTRUMP FV&V report needs to demonstrate the acceptability of the smoothing logic. Response: The assessment cases are included in the NOTRUMP Final Validation Report, WCAP 14807, Revision 1, which has been submitted via Westinghouse letter NSD-NRC 97-4960, dated 1/31/97.
l i i l k l NOTRUMP RELATED ITEMS FOR WHICH WESTINGHOUSE . BELIEVES THE (W) STATUS SHOULD BE CHANGED TO CLOSED IN OITS (3/27/97 Page 4) ' r ITEM 3154 DSER-Ol 21.6.2.4 13 Westinghouse needs to submit the assessment calculations to demonstrate acceptable logic ; operation and logic interactions during the FV&V of the AP600 NOTRUMP code. ; Response: The assessment cases are included in the NOTRUMP Final Validation Report, WCAP 14807, Revision 1, which has been submitted via Westinghouse letter NSD-NRC 4960, dated 1/31/97. ITEM 3155 DSER-Ol 21.6.2.5-1 Westinghouse needs to address the models affecting the fluid entering the ADS piping, ; particularly for the hot legs and pressurizer in the FV&V report. Response: The NOTRUMP Final Validation Report, WCAP 14807, Revision 1, includes comparisons of the test data versus simulation results for the ADS flows. < ITEM 3156 DSER Ol 21.6.2.5-2 - : Westinghouse needs to investigate the NOTRUMP code's inability to properly characterize , CMT thermal stratification and to better explain some of the differences in CMT discharge flow > comparisons. , Response: Section 6 of the NOTRUMP Final Validation Report, WCAP-14807. Revision , ! contains the CMT test simulations which were redone after the preliminary calculations along with a discussion of these tests. ITEM 3160 DSER Ol 21.6.2.6-3 Westinghouse needs to submit reanalysis of the integral systems tests listed in Table 21.10.
)
1 Response: The analyses are included in the NOTRUMP Final Validation Report, WCAP-l 14807, Revision 1, which has been submitted via Westinghouse letter NSD-NRC 97 4960, I dated 1/31/97. ITEM 3161 DSER-Ol 21.6.2.7 1 Westinghouse needs to address PRHR primary side heat transfer comparisons between NOTRUMP and OSU/SPES 2 data in the NOTRUMP Final Validation Report. Response: PRHR heat transfer comparisons were included in Section 7 of the NOTRUMP l Final Validation Report for SPES 2. These comparisons were not included for OSU since i detailed test data were not available to calculate appropriate heat transfer rates for comparison to the NOTRUMP simulations.
t 1 NOTRUMP RELATED ITEMS FOR WHICH WESTINGHOUSE BELIEVES THE (W) STATUS SHOULD BE CHANGED TO CLOSED IN OITS i (3/27/97 Page 5) l. ITEM 3162 DSER Ol 21.6.2.7 2 The NOTRUMP FV&V report needs to address the effects of noncondensible gases on PRHR heat transfer. i 1 Response: The effect of noncondensible gases on PRHR heat transfer is discussed in the ,_ response to RAI 440.325 which is being included in the NOTRUMP Final Validation Report as l part of the RAI appendix.
- 1 i
ITEM 2820 RAl Ol 440.466 (PVR FOR OSU TESTS, LTCT-GSR-001) 1 ~ i The current OITS status is in error. This was completed in the original RAI response and no l
- verification calculations were committed to. This should be changed to
- .
l
" Closed - response provided via Westinghouse letter NTD-NRC 95-4587, dated 11/3/95."
l , f j ITEM 2821 RAl-OI 440.467 (PVR FOR OSU TESTS, LTCT-GSR-001) i The current OITS status is in error. This was completed in the original RAI response and no
' verification calculations were committed to. This should be changed to:
- " Closed - response provided via Westinghouse letter NTD NRC-95-4587, dated 11/3/95."
l f i
, ITEM 2823 RAl-Ol 440.469 (PVR FOR OSU TESTS, LTCT-GSR-001) i '
The OITS should be changed to: " Closed - the NOTRUMP Final NOTRUMP Final Validation
; Report, WCAP-14807, Revision 1, which has been submitted via Westinghouse letter NSD-NRC 97 4960, dated 1/31/97 provides the requested equations in Section 2.4 and the requested benchmark in Section 3.5."
ITEM 2824 RAl-Ol 440.470 (PVR FOR OSU TESTS, LTCT-GSR-001) This RAI asked a number of questions related to the Horizontal Stratified Flow Model used in the preliminary calculations. Use of this model was discontinued after the preliminary calculations were performed, and all calectations included in the NOTRUMP Final Validation Report were done without this model. Therefore, the model description is r ot included in the NOTRUMP Final Validation Report and the RAI no longer applies. The OITS should be changed to: " Closed - after the preliminaly exulations were performed, this model was no longer used. The preliminary calculations wcm redone without this model for inclusion in the NOTRUMP Final Validation Report. Therefore, the model description is not l included in the NOTRUMP Final Validation Report and the RAI no longer applies." j 1
!' j
L ! 'e NOTRUMP RELATED ITEMS FOR WHICH WESTINGHOUSE BELIEVES THE (W) STATUS SHOULD BE CHANGED TO CLOSED IN OITS l (3/27/97 Page 6) i ITEM 2825 RAl Ol 440.471 (PVR FOR OSU TESTS, LTCT-GSR-001) !
- The current OITS status is in error. This was completed in the onginal RAI response and no f verification calculations were committed to. This should be changed to: l
" Closed - response provided via Westinghouse letter NTD NRC 95-4598, dated 11/17/95."
t ITEM 2826 RAl Ol 440.472 (PVR FOR OSU TESTS, LTCT-GSR 001) :
' The current OITS status is in error. This was completed in the original RAI response and no verification calculations were committed to. This should be changed to:
" Closed - response provided via Westinghouse letter NTD NRC-95-4594, dated 11/10/95."
l ITEM 2827 RAl Ol 440.473 (PVR FOR OSU TESTS, LTCT-GSR-001) ! The current OITS status is in error. This was completed in the original RAI response and no i verification calculations were committed to. This should be changed to:
" Closed - response provided via Westinghouse letter NTD-NRC-95-4587, dated 11/3/95."
ITEM 2832 RAl Ol 440.478 (PVR FOR OSU TESTS, LTCT-GSR-001) This RAI pertains to the Birthing Model used in the preliminary calculations. Use of this model ! was discontinued after the preliminary calculations were performed, and calculations included in the NOTRUMP Final Validation Report were done without this model. Therefore, the model > description is not included in the NOTRUMP Final Validation Report and the RAI no longer applies. 4 The OITS should be changed to: " Closed - after the preliminary calculations were performed, this model was no longer used. The preliminary calculations were redone without this model for inclusion in the NOTRUMP Final Validation Report. Therefore, the model description is not included in the NOTRUMP Final Validation Report and the RAI no longer applies." ITEM 2833 - RAl-OI 440.479 (PVR FOR OSU TESTS, LTCT GSR-001) The current OITS status is in error. This was completed in the original RAI response and no ; verification calculations were committed to. This should be changed to:
" Closed - response provided via Westinghouse letter NTD-NRC-96-4626, dated 1/19/96."
i i NOTRUMP RELATED ITEMS FOR WHICH WESTINGHOUSE BELIEVES THE (W) STATUS SHOULD BE CHANGED TO CLOSED IN OITS (3/27/97 Page 7) NOTE REGARDING 440.480 AND 440.481: WESTINGHOUSE HAS COMMITTED TO UPDATE SECTIONS 2.15 AND 2.16 OF THE NOTRUMP FINAL VALIDATION REPORT TO INCLUDE SOME INFORMATION CONTAINED IN THE RESPONSES TO RAIS 440.480 AND 440.481. ITEM 2834 RAl Ol 440.480 (PVR FOR OSU TESTS, LTCT-GSR-001) The current OITS status is in error. This was completed in the original RAI response and no verification calculations were committed to. This should be changed to:
" Closed - response provided via Westinghouse letter NTD NRC-96-4626, dated 1/19/96."
ITEM 2835 RAl Ol 440.481 (PVR FOR OSU TESTS, LTCT-GSR-001) The current OITS status is in error. This was completed in the original RAI response and no verification calculations were committed to. This should be changed to:
" Closed - response provided via Westinghouse letter NTD-NRC-95 4598, dated 11/17/95."
ITEM 2836 RAl Ol 440.482 (PVR FOR OSU TESTS, LTCT GSR-001) The current OITS status is in error. This was completed in the original RAI response and no verification calculations were committed to. This should be changed to:
" Closed response provided via Westinghouse letter NTD-NRC 96-4626, dated 1/19/96."
i I i
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[searrey y. Gpw s., ) Proposed SSAR Sub-Section 15.0.? 15.0.? Loss of Offsite ac Power i As required in GDC 17 of 10 CFR Part 50 Appendix A, anticipated operational occurrences and j postulated accidents are analyzed assuming a loss of offsite ac power. The loss of offsite power is n . j considered as a single failure and the analysis is performed without changing the event category. In i the analyses, the loss of offsite ac power is considered to be a potential consequence of the event. A loss of offsite ac power will be considered a consequence of an event due to disruption of the grid ! following a turbine trip during the event. Event analyses which do not result in a possible consequential disruption of offsite ac power do not assume offsite power is lost. For those events where offsite ac power is lost, an appropriate time delay between turbine trip and the postulated loss of offsite ac power is assumed in the analyses. A time delay of 3 seconds is used, i This time delay is based on the inherent stability of the offsite power grid as discussed in Section 8.2. Following the time delay, the effect of the loss of offsite ac power on plant auxiliary equipment such , as reactor coolant pumps, main feedwater pumps, condenser, startup feedwater pumps and RCCA's is l considered in the analyses. I The AP600 PMS and passive safeguards systems are not dependent on offsite power or on any backup diesel generators. Following a loss of ac power, the PMS and passive safeguards will be able to ; perform the related safety function and there wil! be no additional time delays for these functions to be completed. l I i I (- 1 t l h i 1 i ) i
Application Examples of GDC 17 in AP600 Analyses Imoact Catenories for considerinn GDC imolementation in analyses
- 1) Events initiated with plant at power and connected to the grid (Mode 1)
- 2) Events initiated with plant at HZP (Mode 2) or lower modes
)
- 3) Events initiated from 1.ow power'with plant auxiliaries supplied with ac power by the ,
generator. 1
- 4) At Power events which do not results in a reactor / turbine trip l
- 5) Events which are initiated due to turbine / generator faults l Definitions Attr - Time delay between reaching reactor trip setpoint and turbine trip. Generally, conservative minimum value (0.0 ?) will be assumed unless maximum value is -
demonstrated more severe. i Atn 1,, - Reactor trip signal delay. Time delay from when trip setpoint is reached and , when signal reaches the trip breakers. Includes sensor delay and PMS , processing delay. (~0.2 to ~7.5 seconds) l Atr,, ex , Time Delay for reactor trip breakers to open ( A ten,,,, - Time delay for RCCA grippers to release Atcoc n - Time delay from turbine trip until grid becomes unstable and offsite power is l lost Ato, - For cases where generator is supplying all house loads. Plant is offline (not , connected to grid). This is the time delay between turbine trip and point at which turbine generator inertia can no longer maintain adequate ! voltage / frequency such that RCPs and other auxiliary equipment can still function (up to 30 seconds ?) [ i
s t
- 1) Events initiated with plant at power and connected to the grid (Mode 1) 1A) RCCA Withdrawal at power Time Event i 0.0 RCCA's begin withdrawing i
x Reactor trip setpoint reached (Hi nuclear flux, OTAT) x + Atn Turbine trip occurs : x + Ate,in, RCCA's begin dropping into core
+ Atrn, ex
+ Atan,,,,
x + Atn As a consequence of turbine trip, grid becomes unstable & offsite power is i
+ Atcoei, lost RCP's begin coasting down, feedwater lost IB) Feedline Rupture (initiated at 100% power)
. Time Event 0.0 Feedline rupture occurs x Low steam generator level reactor trip setpoint reached x + Ar n Turbine trip occurs x + Ata, in, RCCA's begin dropping into core
+ atrn, ,x
+ Aten,,,,
1 i x + Atn As a consequence of turbine trip, grid becomes unstable & offsite power is
+ Atcoc i, lost RCP's begin coasting down, feedwater lost
i
- 2) Events initiated with plant at HZP (Mode 2) or lower modes
- Turbine is offline all ac power is supplied from offsite ac power 2A) RCCA Withdrawal from Suberitical Conditions Time Event l 0.0 RCCA's begin withdrawing x Hi flux reactor trip serpoint reached 1-2>,, T u d ! = : rip c c =
x + Ata,vn, RCCA's begin dropping into core
+ At t,, ,x :
+ Ata,,,, l x + Atrr As a consequence of turbine trip, grid becomes unstable & offsite power is j
+ Atcoe n lost :
1 RCP's begin coasting down, feedwater lost j Plant reactor trip does not cause a consequential grid disruption. Offsite ac power continues to ) supply auxiliary loads such as RCP's. ; l 1 I l J l l l
- I l
I I i I i l
- 3) Events initiated from low power with plant auxiliaries supplied with ac power by the generator.
Offsite ac power is not connected to plant Plant auxiliaries (RCP's, feedwater pumps, etc) are supplied by generator l 3A) RCCA Withdrawal at power (low power < ~ 10%) i Time Event 0.0 RCCA's begin withdrawing f x Reactor trip setpoint reached (Hi nuclear flux) l t x + atn Turbine trip occurs. Turbine generator begins coasting down. ' x + At ,1,, RCCA's begin dropping into core
+ Attr, ex
+ Ato, x + Atn - Turbine / generator reaches speed such that voltage / frequency decrease such l
+ Atc that RCP's can not continue to function. RCP's begin coasting down. !
. I i
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- 4) At Power events which do not results in a reactor / turbine trip j Due to setpoint uncertainties, PMS channel / division out of service (as permitted by Tech Specs), failure of PMS channel / division, some analyses do not assume protection function occurs. Examples are excessive load increase and inadvertent PRHR actuation while at power. l These events ride out the transient and reach an equilibrium condition without reactor l trip. l
- Review affected SSAR events & assume all PMS channels / division operate and/or that !
setpoint uncenainties reactor trip & turbine trip occur. 4A) Current SSAR inadvenent PRHR actuation (without reactor trip) .l I l Time Event t 0.0 PRHR actuated with plant at 100% power causing an asymmetric power i excursion. , x, Overpower trip setpoint (hi flux or OPAT) reached on loop with PRHR. ! One PMS channel on PRHR loop is assumed out of service. One PMS ! channel on PRHR loop is assumed to fail. Maximum setpoint uncenainties are assumed such that trip channels on ! loop opposite the PRHR do not reach reactor trip setpoint. ; 2/4 trip logic is not satisfied and no reactor trip occurs f I x2 Peak Core power reached : i x3 Core power reaches equilibrium with steam generator and PRHR heat ! removal rate. i 4B) AdditionalInadvenent PRHR Actuation (with reactor trip & loss of ac power) 1 (To Be Supplied) { l i I I l j
),
L l
- 5) Events which are initiated due to turbine / generator faults !
(To Be Supplied) i i W f a s t I l i e
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** TX CONFIRMATION REPORT **
AS OF APR 3 '97 13:46 PAGE.01 s APG00 DESIGN CERT : l DATE TIME TO/FROM MODE MIN /SEC PGS STATUS 01 4/ 3 13:44 #23:NRC G3--S 01'49 03 OK Post-it' Fax Note 7671 oawg[3[9 7 l4> 3 T* T~ee Qu A} *** fa t &crM s
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, PHONE: FACSIMILE: PHONE: Office:
COMPANY: Facsimile: win: 284-4887 jV[4 6 outside: (412)374-4887 LOCATION: Rocj<vij_L[f l l Cover + Pages 1+$ i . The following pages are being sent from the Westinghouse Energy Center, East Tower, , Monroeville, PA. If any problems occur during this transmission, please call: ! WIN: 284 5125 (Janice) or Outside: (412)374 5125. i COMMENTS: A TTAcHEh AA E SsArt MAAu u?s
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- 7. Instrumentation and Controts The potection and safety monitoring system includes the following:
Integrated protection cabinets
. Engineered safety features actuation cabinets Protection logic cabinets I
- Qualified data processing cabinets 1
- Qualified data processing I/O cabinets I
- Qualified displays Reactor trip switchgear
- Sensors Main control room and remote shutdown workstation multiplexers
- p mom Special cent.r.1 Monitoring room / remets skutc)euan woorlttfobon fransfer pone /5 System ne special monitoring system does siot perform any safety-related or defense-in-depth functions. The special monitoring system consists of specialized subsystems that interface with the instrumentation and control architecture to provide; diagnostic and long term monitoring functions.
4 He special monitoring system includes the metal impact monitoring system. The metal impact monitoring system detects the presence of metallic debris in the reactor coolant system
- when the debris impacts against the intemal pans of the reactor coolant system. He metal impact monitoring system includes digital circuit boards, controls, indicators, power supplies I
and remotely located sensors and related signal processing devices. The sensors and their related signal processing devices are mounted in pairs to maintain the impact monitoring I function if a sensor fails in service. De metal impact monitoring system is desenbed in subsection 4.4.6.4. Plant Control System he plant control system provides the functions necessary for normal operation of the plant from cold shutdown through full power. He plant control system controls nonsafety related components 'in the plant that are operated from the main control room or remote' shutdown workstation. De plant control system contains nonsafety related control and instrumentation equipment to change reactor power, control pressurizer pressure and level, control feedwater flow, and perform other plant functions associated with power generation. The plant control system includes the following: Distributed controllers Signal selectors Rod contro! cabinets Rod drive motor generator sets Pressunzer heater control interface Revision: 5 g February 29,1996 7.1 4 y Westingh0use
- 7. Instrumentation and Controls i
1 The I/O cabinets are microprocessor based, safety-related modular data gathering units. The 1/O cabinet can receive inputs from process sensors and safety related digital systems. He 1/O cabinet consolidates the input data, performs conversions to process units, and formats the I l data for the data link transmission to the qualified data processing cabinets. 7.1.2.7 Main Control Room and Remote Shutdown Workstation Multiplexers ne protection and safety monitoring system contains eight multiplexers. One main control ! room multiplexer and one remote shutdown workstation multiplexer is associated with each of the four safety divisions. Each multiplexer consists of two redundant halves or subsystems. I The multiplexers provide for transmission of component level manual actuation signals from the main control room or remote shutdown workstation to the protection logic cabinets. The multiplexers also provide for transmission of component status information from the protection logic cabinet to the main control room and remote shutdown workstation. 1 The multiplexers communicate with soft control devices or operator interface modules in the main control room or the remote shutdown workstation over redundant fiber-optic data links. Subsection 7.1.3.4 provides additional discussion of the operation of the soft control devices. Vanous " handshaking" signals are implemented for requests and responses between the soft controls and the multiplexers to verify the receipt and the validity of the messages. -
~
7.1.2.8 Sensors Yc brans [ee er( cae f rol Som St '" * *
- C oh r* I * **
rem.g u d sw,k a s a,s ejeserned %,pi.e.on u occomp),A,d using bors[er l
;, siseeban 7.4 3. .
The protection and safety monitoring system monitors key variables related to equipment mechanical limitations, and variables directly affecting the heat transfer capability of the reactor. Some limits, such as the overtemperature AT serpoint, are calculated in the integrated protection cabinets from other parameters because direct rneasurement of the variable is not possible. His subsection provides a description of the sensors which monitor the vanables for the protection and safety monitoring system. For convenience the discussions are grouped mto the following three categories: Process sensors Nuclear instrumentation detectors Status inputs from field equipment I The inputs described are those required to generate the initiation signals for the protective I functions. The use of each parameter is discussed in the sections that deal with each I protective function. For example, reactor trip is discussed in Section 7.2 and engineered safety features actuation is desenbed in Section 7.3. 7.1.2.8.1 Process Sensors The process sensors are devices which measure temperature, pressure, fluid flow, arid fluid I level. Process instrumentation excludes nuclear and radiation measurements. Revision: 5 s February 29,1996 7.1-20 T Westingh0054
- 7. I strumentatits rrd Centrols I
EC 880-1986; " Software for Computers in the Safety Systems for Nuclear Power
; Generating Stations" EEE 828-1983; "EEE Standard for Software Configuration Management Plans" EEE 8291983; "EEE Standard for Software Test Documentation" EEE 830-1984; "EEE Standard for Software Requirements Specifications" l
- EEE 1012-1986; "EEE Standard for Software Verification and Validation Plans"
, I
- EEE 1042-1987; "EEE Guide to Software Configuration Management (ANS0" WCAP-13383 also provides for the use of commercial off-the-shelf hardware and softwar through a commercial grade dedication process.
- _f 7.1.3 Plant Control System 4
- QGVL56 As Shou)U W IMs CA T 7. I.2 M-!
' The plant control system is a nonsafety related system that provides control and coordination
. of the plant during startup, ascent to power, power operation, and shutdown conditions. He
. plant control system integrates the automatic and manual control of the reactor, reactor coolant, and various reactor support processes for required normal and off normal conditions.
The plant control system also provides control of the nonsafety-related decay heat removal
' systems during shutdown. The plant control system accomplishes these functions through use of the folloving:
- Rod control
- Pressurizer pressure and level control
- Steam generator water level control Steam dump (turbine bypass) control
- Rapid power reduction The plant control system provides automatic regulation of reactor and other key system parameters in response to changes in operating limits (load changes). The plant control i system acts to maximize margins to plant safety limits and maximize the plant transient i performance. The plant control system also provides the capability for manual control of l plant systems and equipment. Raduad=at control logic is used in some applications to increase single failure tolerance.
The plant control system includes the equipment from the process sensor input circuitry through to the modulating and nonmodulating control outputs as well as the digital signals to other plant systems. Modulating control devices include valve positioners, pump speed controllers, and the control rod equipment. Nonmodulating devices include motor starters for motor-operated valves and pumps, breakers for heaters, and solenoids for actuation of air-operated valves. The con'rol cabinets contain the process sensor inputs and the modulating Revision: 11 7.1 27 February 28,1997 3 %ikiiiGUSS
- 7. Instrumentation annt Controls standards, and practices aimed at maximizing reliability and safety. For example, wiring used within electrical equipment, and devices used to protect winng from overcurrent (such as
- breakers, fuses. and current limiters), are sized and coordinated according to National Electric Code. Insulation used is flame retardant and meets National Electnc Code. IEEE and Underwriter's Laboratory guidelines applicable to the environment where the winng is located. Electronics are housed in cabinets of metal construction. Isolation devices are incorporated into wiring leaving the protection cabinets to the other redundant protection cabinets or nonsafety-related areas The independence of electrical equipment is verified as discussed in subsection 7.1.4.2.6. L de);[,on de &c speben peledol is ll-sh) Lf //e /%
'7.1.4.2.6 pome- levej of Os eksyntol e85"en1!" C'd b5 +h e U.Sc Conformance to the Requirements to Maintain Channel Independence (Paragraph 4.6 of of IEEE 2791971, GDC 22, IEEE 384-1974, Regulatory Gulde 1.75) 6 bee The flexibility of the protection and safery monitoring system enables physical sepa redundant divisions.
Where redundant equipment communicates, such as at the integrated protection cabinets, isolation devices are employed to preserve electrical independence of the divisions. These
' devices are desenbed in subsection 7.1.2.11. They are also used to preserve the independence of safety equipment from nonsafety related systems which may use protection signals.
Nonsafety-related wiring is separated from safety-related wiring as discussed in Chapter 8. Analyses, tests, or physical barriers are used to verify the adequacy'of wire routing where separation distances are less than those suggested by regulatory guides or industry standards. 4 He physical separation criteria for protection system cabinets includes the applicable , recommendations contained in Paragrapls 6 6 of IEEE 384 (Reference 6). Specific criteria
- applied are the following
- 1 Internal separation criteria pertaining to separation between redundant Class IE equiprnent according to Paragraph 6.2 of IEEE 384-81 1
= 1 Non-Class IE wiring criteria pertaining to separation between Class IE wiring and non-Class IE wiring according to Subparagraph 6.6.5 of IEEE 384-81
=
Cable entrance criteria of redundant Class IE cables ac:ording to Subparagraph 6.6.6 of IEEE 384-81 l The application of these criteria to instrumentation cabinets is endorsed by Regulatory Guide 1.75. 4 Wirin's for redundant divisions use physical separation, analyses, isolation, tests, or barriers to provide independence of the circuits. 1 g 1 Revision: 5 y Westinghouse 7.1-41 February 29, 1996
.. - .._-_ . -_ . - . . ~ . . . - - . _ . ..
SSAR INSERT 7.1.2.15-1 WCAP-13383 provides a planned design process for hardware and software development during the following life cycle stages:
- Design requirements phase j
= System definition phase
- Hardware and software development phase
- System test phase ,
e installation phase WCAP-13383 also provides for the use of commercial off-the-shelf hardware and software through a commercial dedication process. Control of the hardware and ' software during the operational and maintenance phase is the responsibility of the Combined License applicant as described in Subsection 13.5.1. 1 i 1 1 l 1 6
O e I l / __ Wo Toa.uS di il ib WAN REWCrit CONTR(1 Sr4m hiF'E0 Ss0NAL DEDCA N * " " ' " OW"I SELECTION
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, l WATIPLEXERS
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- 7. Instrumentatica and Controls 4
Automatic Actuadon Function The automatic actuation signals provided by the diverse actuation system are generated in a functionally diverse manner from the protection system actuation signals. The common-mode failure of sensors of a similar design is also considered in the selection of these functions. The automatic actuation function is accomplished by redundant microprocessor based subsystems. Input signals are received from the sensors by an input signal conditioning block, which consists of one or more electronic modules. This block converts the signals to standardized levels, provides a barrier against electromagnetic and radio frequency interference, and presents the resulting signal to the input signal conversion block. The conversion block continuously performs analog to digital signal conversions and stores the value for use by the signal processing block. i j The signal processing block polls the various inputs under the control of a software-based algorithm, evaluates the input signals against stored serpoints, executes the programmed logic when thresholds are exceeded, and issues actuation commands.
- The resulting output signals are passed to the output signal conversion block, whose function l is to convert microprocessor logic states to parallel, low-level de signals. These signals are passed to the output signal conditioning block. This block provides high level signals capable of switching the traditional power plant loads, such as breakers and motor controls.* lt also a provides a barrier against electromagnetic and radio frequency interference.
4 Diversity is achieved by the use of a different architecture, different hardware implementations 1 and different software from that of the protection and safety monitoring system. The diverse design uses standard input modules designed for use with small industrial computer systems. It also uses a microprocessor board different from those used in the
- protection system. I Software divenity is achieved by ruiming different operating systems and programming in different languages.
, The diverse automatic actuations are:
Trip rods via the motor generator set, trip turbine and initiate the passive residual heat removal on low wide range steam generator water level 4 Initiate passive residual heat removal on high hot leg temperature Trip rods via the motor generator set, actuate the core makeup tanks, and trip the reactor coolant pumps on low pressurizer water level scleele.cl } Isolate aneseppontainment penetrations and start passive containment cooling water flow j on high containment temperature Revision: 5 February 29,1996 7,7 16 [ Westinghoust
o
, 7 Intrumentation cod Controts
% u) kith Orc, ISofo(ec) 13Ne diver 2 ockoodeon A Gnheel containment penetration are those lines that connect directly to the reactor coolant 93 ^'
system, the contamment atmosphere, or the containment sump l The selection of setpomts and time responses determine that the automatic funcuons do not actuate unless the protecnon and safety momtonng system has failed to actuate to control plant conditions. Capabilirv is provided for testmg and calibratmg the channels of the diverse actuation system. ! Manual Actuation Function "the manual actuation function of the diverse actuation system is implemented by winng the controls located in the mam control room directly to the final loads in a way that completely bypasses the normal path through the control room muluplexers, the engineered safety features actuation cabinets, the protecuon logic cabinets and the diverse actuation system automauc , lope. e open she I shm {h Ap,,,mg The diverse manual functions are: Sytiem volvea
- Reactor and turbine inp
' '88d'irob
' * * "3 i
- Passive residual heat removal actuation l
- Core makeup tank actuation I* 8E*$t 3 owlemise dviruve,W j m...._ c.z::": :, :^:- +: r":: 818Iem vof v,3 l
-
- Passive containment cooling actuauon N ,opea s b e 9 o g , " '4 j 'e " 15 u 2*le l s
Scledeel e* Ctmeal containment contamment penetration hydrogen igniier actuationisolation
" des j
- Initiate in containment refueling water storage tank injection ,
I
- Initiate contamment recirculation I l
- Ininate in-contamment refueling water storage tank drain to containment I
- Actuation Logic Function There are two actuation logic modes, automauc and manual. The automatic actuation logic ;
l l mode functions to logically combine the automanc signals from the two redundant automatic
- subsystems in a two-out-of two basts. The combined signal operates a power switch with an output drive capability that is compatible, in voltage and current capacity, with the i requuements of the final actuation devices. The tworout of two logic is implemented by connecting the outputs in series. The manual actuation mode operates in parallel to a independently actuate the final devices.
j Actuation signals are output to the loads in the form of normally de energized, energize to-
- I actuate signals. The normally de energtzed output state, along with the dual, two out of two
] redundancy reduces the probabtlity of inadver:ent actuation. The diverse actuauon system is designed so that, once actuated, each mitigation action goes to complenon. Any subsequent retum to operanon requires deliberate operator action. 4 Revision: 10 7.7 17 December 20,1996 g 3 WO6Wighoust
[WW) WestinghouseFAX COVER SHEET RECIPIENT INFORMATION SENDER INFORMATION DATE: /Qpe,c 9 ,997 NAME: L gg, I TO: LOCATION: ENERGY CENTER - beAcc 1Acecc4 EAST PHONE: FACSIMILE: PHONE: _off;ce: d/2-374-r29o COMPANY: Facsimite: win: 284 4887 US Ald c . outside: (412)374 4887 LOCATION: a Cover + Pages 1+f The fo!!owing pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call: WIN: 284-5125 (Janice) or Outside: (412)374 5125. COMMENTS: D tAar He,< ,s t k n k s +l J ckeald ec,cive Clucu a nc,,, en Om D s l oce. sus Clw E' O)e cbs ssse d PL s ceree<n </e d e d w an eu r { k, ke (lall.
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,, 3. Design of Structures yomponents, Eq11pment and Systems For Class D structures, systems, and components containing radioactivity, it is demonstrated by conservative analysis that the potential for failure due to a design basis event does not result in exceeding the normal offsite doses per 10 CFR 20. This criterion is in conformance with the definition of Class D in Regulatory Guide 1.26. -Seme nrect=:, :y;:cm; and #
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- A structure, system or component is classified as Class D when it directly acts to prevent unnecessary actuation of the passive safety systems. Structures, systems and components which support those which directly act to prevent the actuation of passive safety systems are also Class D. The inclusion of these nonsafety-related structures, systems, and components in Class D recognizes that these systems provide an important first level of defense that helps to reduce the calculated probabilistic risk assessment core melt frequency. These structures, systems, and components are normally used to support plant cooldown and depressurization and to maintain shutdown conditions during maintenance and refueling outages.
For Class D structures, systems, and components considered to be risk significant as defined in the reliability assurance plan (see Section 16.2). Provisions are made to check for operability, including appropriate testing and inspection, and to repair out-of-service stmettites, systems, and components. These provisions are documented and administered in the plant reliability assurance plan and operating and maintenance procedures. Some Class D ctructures, systems, and components are assumed to function in a severe containment environment. The design requirements for these components include operation in such an environment. An evaluation is done to confirm that the structure, system, or component can be expected to function in such an environment. Standard industrial quality assurance standards are applied to Class D structures, systems, and components to provide appropriate integrity and function although 10 CFR 50, Appendix B and 10 CFR 21 do not apply.10 CFR 50, Appendix B and 10 CFR 21 do apply to Class D structures, systems, and components that are seismic Category I. These industrial quality assurance standards are consistent with the guidelines for NRC Quality Group D. The industry standards used for Class D structures, systems and components are widely used industry standards. Typicalindustrial standards used for Class D systems and components are provided as follows:
- Pressure vessels - ASME Code, Section VIII
- Piping - ANSI B 31.1. Power Piping, (Reference 5)
= Pumps - AIT610 (Reference 6), or Hydraulic Institute StandarA (Reference 7)
Valves - ANSI B16.34 (Reference 8) Revision: 11 Y W85tingh0088 3.2-9 February 28,1997 4
d-e 4
** TX CONFIRMATION REPORT ** AS OF APR 9 '97 15:03 PAGE.01 APSOO DESIGN CERT DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 4/ 9 15:02 #23:NRC G3--S Ol*09 02 OK l
t i i l y l Brian A. McIntyre,06:25 PM 4/8/97 ,4/3 RTNSS meeting l Date: Tue,8 Apr 199718:25:46 0400 ) To: TRQ@NRC. GOV From: " Brian A. McIntyre" <mcintyba@wesmail.com> j
Subject:
4/3 RTNSS meeting ' Cc: haagcl@ wesmail.com, meintyba@ wesmail.com Ted, Please pass this to Bill Huffman. Bill, My notes from the 4/4 RTNSS meeting indicate that the following three things will happen:
- 1. Westinghouse will examine the material discussed to see how these ideas cocid be used to provide additional margin between the focused PRA and the baseline PRA such that the thermal hydraulic uncertainty effort can be cased. We will also look at what could be done to provide additional margin for long term cooling.
- 2. The NRC will determine how their position of adding additional regulatory oversight to a nonsafety system would answer their questions that have been asked on thermal hydraulic uncertainty. This will include -
determining what the concept of margin merns in long term cooling (the regulatory requirement of 10 CFR 50.46(b)(5) is that the " calculated core temperature shall be maintained at an accepetably low value and decay heat shall be removed for the extended long period of time .." so is it feet of water over the top of the core?? If the core is covered, decay heat must be being removed)
- 3. 'Ihe NRC staff will review the regulatory oversight proposed by ,;
Westinghouse in WCAP 13856, September 1993 to determine ifit is ' appropriate, given the RTNSS mission of the system. Since the post-72 hour issue is a subset of the RTNSS issue, this review should include the proposed additional regulatory oversight included in our March 14,1997 letter on this subject. If you disagree with these items, please let me know. We need to be ready 1 to say something on April 18. Cindy and I are talking about how to respond to our item and will provide you'with a date by 4/11. l Printed for " Brian A. McIntyre" <mcintyba@wesmail.com> 1l
I. l
.i When will we hear from you with your either date of results??
Brian A.McIntyre 412.374.4334 WIN 284.4334 i FAX 374.4887 i l i t
)
l i s l t i l r l Printed for " Brian A. McIntyre" oncintyba@wannail.com> 2 J. l 4
l QuayNRC,04:05 PM 4/6/97, Request for letter j To: QuayNRC From: " Brian A. McIntyre" <mcintyba@wesmail.com>
Subject:
Request for letter Cc: BAM Bec: X-Attachments:
- Ted, In Tim's letter of March 27,1997 concerning Key Issue #1, he mentions a letter of January 7,1997 where the staff identified key issues that could become critical path in the AP600 review. I have looked in my files and while I have the NRC December 6,19% letter, I cannot find my copy of the NRC January 7,1997 letter.
I would appreciate it if you would fax me a copy of the January 7 letter Monday moming so that I may accurately complete my response to Tim's letter. Thanks!!! l Printed for " Brian A. McIntyre" <mcintyha@wesmall.com> 1l l I J
1
% Cc W ? ns f k( -> _ e 'SG7 C O V E R j\ e Brian A. Mc Intyre S H E E T To: Bill Huffman
Subject:
Today's Telecon Date: April 7,1997 Pages: Three, including this cover sheet. COMhENTS: Bill, Attached is DSER open item and proposed wording for OITS item. Let's discuss at lpm today. Thanks. cc: B. McIntyre (NRC Informal File), L.Hochreiter, From the desk of.. R.Kemper, R.Osterrieder, File 7.6 Eari H. Novendstem Manager, Advanced and WER Plant Safety Analysis Westinghouse Po Box 355 PRtsburgh, PA 15235 (412) 374 -4790 Fax: (412) 374-4011 April 7, 1997 I:\NRC\TELECON.497 I
7 7
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DSER CONFIRMATORY ITEM (Q11, n ").h NNf g DSER CN 21.6.2.4-4 Westinghouse needs to venfy that heat link methodology for transition ooiling if, not used in AP600 NOTRUMP calculations.
Response
Based upon the wording of this question and the discussion on page 21123 of Reference 1, there may be a misunderstanding of the implementation and/or application of the revised colution scheme for transition boiling in the noncritical heatlink calculations of NOTRUMP for AP600. The followir:g clarification is provided. The Westinghouse transition boiling correlation is used in two separate models o' @ NOTRUMP code for post DNB heat transfer. One is in the non-critical heatlink calculations, which are not used for modelinE core heat transfer. The other is in the fuel rod-to-fluid heat transfer coefficient calculations (between core nodes and fluid nodes). These are discussed in more detail below. The Westinghouse transition boiling correlation as implemented in the non-critical heatlink calculations of NOTRUMP is discussed in Section 6 of Reference 2. The wall temperature calculations via Equation 6-40 of Reference 2 are performed with an iterative solution technique. The iterative scheme which was originally employed was a crude method of successive substitution. Recently, this was improved by replacing the original scheme with the half-interval method which guarantees convergence, as discussed in Section 2-19 of the NOTRUMP Final Validation Repon for AP600 (Reference 3) . This improvement is used for all non critical heatlinks, and thus it is used in all of the AP600 analyses documented in Reference 3 in which non-critical heatlinks are employed. The Westinghouse transition boiling correlation as implemented in the fuel rod-to-fluid heat transfer coefficient I calculations (between core nodes and fluid nodes) of NOTRUMP is described in Appendix T of Reference 2. specifically in Section T-3 via Equations T-25 and T-26. 'Ihere have been no modifications in this area of the code, and thus there are no changes to the core heat transfer methodology employed in the AP600 analyses of Reference 3. References
- 1. " Supplement to the Draft Safety Evaluation Report: Related to the Cenification of the AP600 Design" U. S. Nuclear Regulatory Commission, Docket No. 52-003, April 1996.
- 2. Meyer P. E., et. al., "NOTRUMP - A Nodal Transient Small-Break and General Network Code," Westinghouse Electric Corporation, WCAP-10079-P-A (Proprietary), WCAP-10080-P-A (Non-Proprietary), August 1985.
- 3. Fittante R. L., et. al., "NOTRUMP Final Validation Repon for AP600," Westinghouse Electric Corporation, WCAP-14807 Revision I, January 1997,
m . . . _ . . ,_ m - m . . . _ . . _ _ - i (/, - h y , c j f$ j l W $ b Y M M <. 4 0 % m * * / > a u g p o k ee- n o L 4 A p s' 1 xacew -
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. t C O V E R S H E E T FAX i To:
Subject:
Bill Huffman Loss of Off-Site Power - ' {9 L s Date: April 3,1997 APR - 71997 Pages: Three, including this cover sheet. f_.1 ( C Brian A. Mc Intyre COMMENTS: Bill, l Attached is explanation of why the gird is stable for three seconds. Once you and Summer give us feedback, we will issue in separate letter to NRC. Let's discuss Wednesday. Also, if you have technical questions, let me know before our phone call whether we need Tom Hayes, our electrical expen, to be available. Thanks. i i, E i i cc: B. McIntyre (NRC Informal File), E. Carlin, R. Nydes, T. Hayes, T. Schulz, File 7.6 From the desk of... Earl H. Novendstern Manager, Advanced and VVER Plant Safety Analysis Westinghouse PO Box 355 Pittsburgh, PA 15235 s (412)374 4790 Fax: (412) 374 4011 Apr61 3. 1997 g entc x Luap.1 4 99
_ _ .- - ._ - - - _ .. - , - . ~ .. . - . . - l? [ l.d l 1 AP600 GRID STABILITY REQUIREMENTS l l The AP600 is designed with passive safety-related systems for core cooling and containment integrity and, therefore, does not depend on the electric power grid for safe operation. This i feature of the AP600 significantly reduces the importance of the grid connection and the ' requirement for grid stability, The AP600 safety analyses assume that the reactor coolant pumps (RCPs) can receive power from either the main generator or the grid for approximately 3 seconds following a turbine trip. Normally, The RCP will receive power from the grid (if the grid is connected and stable). If the grid is disconnected prior to the turbine trip this RCP power will come from the stored energy in the inertia of the main turbine-generator. There is a possibility that an unstable grid (with improperly functioning protective devices) could collapse as a consequence of the turbine trip and "take" the inertial energy from the main turbine-generator before the required 3 seconds. The probability of this event is immeasurably small for the following reasons: ) l 1. As stated in Branch Technical Position ICSB-11 (PSB) (attached] an interconnected grid (e.g., the U.S. mainland) meets the grid availability requirements of a conventional i (non-passive) nuclear power plant. There is also a strong indication that an isolated j system large enough to justify inclusion of a nuclear unit will also meet this criterion, , however, additional measures may be needed. This conclusion can be applied with I margin to the AP600 since the AP600 grid requirements are less than those of a ! conventional plant.
- 2. The time-constant associated with an unstable grid will generally be significantly longer than 3 seconds. A time-constant as low as 3 seconds would generally only occur with t island operation where the AP600 is the only unit operating in the island. The '
probability of islanding of a well-designed grid, within 3 seconds of a turbine trip, is ! l extremely small.
- 3. The COL will be required to perform a grid stability analysis to show the grid will remain stable for at least 3 seconds following a turbine trip.
i s l- l l l l L hayesioffsitoistab_3 wpf ' Apnl3.1997 i^ l , _ . . _ .__ ._ _ _ _ _
f l a. i . BRANCH TECHNICAL POSITION ICSB-11 (PSB) STABILITY OF 0FFSITE POWER SYSTEMS A. BACKGROUNO The staff has traditionally required each applicant to perform stability l i studies for the electrical transmission grid which would be used to provide the offsite power sources to the plant. The basic requirement is that loss of ! the largest operating unit on the grid will not result in loss of grid stability and availability of offsite power to the plant under consideration. In some cases, such as plants on the island of Puerto Rico, the plant is connected to an isolated power system of limited generating capacity. These kinds of isolated power systems are inherently less stable than equivalent systems with
- supporting grid interties. It is also obvious that limited systems are more l vulnerable to natural disasters such as tornadoes or hurricanes.
i B. BRANCH TECHNICAL POSITION 1
- 1. The staff has concluded, from a review of appropriate reliability data, that power systems with supporting grid interties meet the grid availabil-ity criterion with some margin. This conclusion is applicable to the
- review of most plants located on the U.S. mainland.
4 2. There is also strong indication that an isolated system large enough to justify inclusion of a nuclear unit will also meet this criterion. However, as a conservative approach, the staff will examine the available generating capacity of a system, including interties if available, to ' withstand outage of the largest unit. If the available capacity is judged marginal to provide adequate stability of the grid, additional measures should be taken. These may include provisions for additional capability and margin for the onsite power system beyond the normal requirements, or other measures as may be appropriate in a particular 4 case. The additional measures to be taken should be determined on an 4 individual case basis. C. REFERENCES None. i 1
;i t.-
Rev. 2 - July 1981 8A-7
o
*v 1
C O V E R S H E E T FAX J To: Bill Huffman
Subject:
Loss of Off-Site Power t -- Q Date: Pages: April 3.1997 Three, including this cover sheet. s \<3 pa 7jgg7 COMMENTS: U'\q C Brian A. Mc Intyre Bill, 4 Attached are " draft" revisions to Chapter 8 and Table 1.8 of the SAR that reflect the three second j delay we will use for Chapter 15 accidents. Please give a copy to Summer Sun and we can
, discuss next Wednesday. Thanks.
Also, I will be sending before our next meeting a write-up on the justification for three seconds. j Once you review this write-up, and we resolve any comments on it, we'll formalize it in a letter to 4 you. I 4 4
=
4 4 4 From tne desk of... Earl H. Novendetern I cc: B. McIntyre (NRC Informal File), E. Carlin, Man *9*r Advanced and WER Plant Sa i R. Nydes, T. Hayes, T. Schulz, File 7.6 wesungnouse PO *iox 355 Pittsburgh. PA .!235 (412) 374 -4790 Fax: (412) 374-4011 4 April ). 1997 1 <NRC\ LOOP 491
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~
- 8. Electric Power 8.2.1.2 Transformer Area The transformer area contains the main stepup transformers, the unit auxiliary transformers.
and the reserve auxiliary transformer. Protectise relaying and metering required for this l equipment is located in the turbine building. The necessary power sources i480 Vac. ; 120 Vac, and 125 Vde) to the equipment are supplied from the turbine building. See subsection 9.5.1 for a discussion of fire protection associated with plant transformers. One feeder connects the transformer area with the switchyard to supply power to/from the main stepup transformers for the unit. An arrangement is shown in Figure 8.31-1. 8.2.2 Conformance to Criteria The offsite sources are not Class IE. Commercial equipment is manufactured to the industrial standards listed in subsection 8.2.5. The design meets General Design Criterion I. Unit trips occur at the generator breaker and do not cause the loss of the preferred power source to the plant electrical systems. The AP600 design does not require ac power for mitigating design basis events *The AP600 meets the intent of General design Criteria 17 as outlined in Section 3.1. % [}nyter /S,0 a escrdes fl>c k 9n buts 41suerffous l pfaler/ A e s./ysa f one c ants Conformance with General Design Criterion 18 is provided by the test and inspection j capability of the system. 8.2.3 Standards and Guides in addition to the General Design Criteria. the industry guides and standards listed as Reference 2 through 4 are used as guides in the design and procurement of the offsite power system. 8.2.4 Combined License Information for Offsite Electrical Power a Combined License applicants referencing the AP600 certified design will address the design of the ac power transmission system and its testing and inspection plan. The Combined License applicant will address the technical interfaces for this nonsafety-related system listed in Table 1.8-1. These technical interfaces include those for ac power requirements from offsite and the analysis of the offsite transmission system. 8.2.5 References
- 1. ANSI C21990. National Electric Safety Code.
- 2. ANSI C37.010-1972. Application Guide for ac High Voltage Circuit Breakers.
l I l Revision: 7 April 30,1996 8.2 2 [ Westiligt100S8
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~
- 1. Introducfon and General Description of Plant Table 1.8-1 (Sheet 3 of 7)
SUhtMARY OF AP600 PLANT INTERFACES
%TTH REMAINDER OF PLANT Matching Section item Interface or Sub.
No. Interface Interface Type item section 6.1 Inservice Inspection requirements for the Requirement of Combined License 6.2.1 containment AP600 applicant program 6.2 Off site environmental conditions assumed AP600 Interface Site specific 6.4 . for Main Control Room and technical parameter support center habitability design . 7.1 Listing of all design criteria applied to the Not an Interface N/A 7 design of the I&C systems l 7.2 Power required for site service water NNS and Not an N/A 7 instrumentation Interface 7.3 Other provisions for site service water NNS and Not an N/A 7 instrumentation Interface , 8.1 Listing of design criteria applied to the NNS Combined License 8 design of the offsite power system applicant coordination 8.2 Offsite ac requirements NNS Combined License 8
! Steady-state load applicant Inrush kVA for motors coordination Nominal voltage Allowable voltage 4
regulation Nominal frequency Allowable frequency fluctuation Maximum frequency decay rate Limiting under frequency value for RCP l 8.3 Offsite transmission system analysis: NNS Combined License 8.2 Loss of AP600 or largest unit applicant analysis Voltage operating range Transient stability n;v37 k' m ain feon e/ 6e i mimasem a f ibret (s.) seronh e f/er. n brbJu 7,;p , 7 II Revision: 9 1.8-5 August 9,1996 3 WBStk1gh0083
y , __. _ 4-u-97 To: Aill //0ffman From : 20hio A>ycbs 1 cc : ren Chi)D7'/JermdfSuggs btn , Jim kWk(s, Eda ' 5vb y ec/: Aak 6kam Lme Leakage 72ch Spec l 8ill, ihre is the redraft af tin subject Tech pec. Akk ,he Emf nods "&cd " cod f+af 14 asse this is sadeguakne may{cr.stico //a's reviewchang heloce r submit & Gemally. % fzynat xhmithd will be win 11, &al recA ye.s end 0f Play. A ye can see, w- dadhd nd to bdlice wth an theanalysis to
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4
I ** Main Steam Line Leakage l B 3.7.8 1
.,-. 1 3.7 PLANT SYSTEMS .
1 3.7.8 Main Steam Line Leakage ! o i l - LC0 3.7.8 Main Steam Line leakage through the pipe walls inside ! containment shall be limited to 0.5 gpm. r 1 i APPLICABILITY: MODES 1, 2, 3 and 4. i ACTIONS (' l CONDITION REQUIRED ACTION COMPLETION TIME ; l A. Main Steam Line A.1 Be in MODE 3. / hours leakage exceeds l operational limit. AND l 1 - A.2 Be in MODE 5.
@ hours j
.\
SURVEILLANCE REQUIREMENTS ! SURVEILLANCE FREQUENCY ! l 1 SR 3.7.8.1 Verify main steam line leakage into the Per SR 3.4.8.1 containment sump < 0.5 gpm. F i i i h AP600 en-s334Dw .spf.14040797 3.7 17 l l
,- 1 l
,.m - . . - -, , . , - -
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' Main Steam Line Leakage !
B 3.7.8 ' s 3.7 PLANT SYSTEMS ;
- 3.7.8 Main Steam Line Leakage j
- i. BASES-2-
BACKGROUND- A limit on leakage from the main steam line inside ! containment is required to limit system operation in the - presence of excessive leakage. Leakage is limited to an amount which would not compromise safety consistent with the Leak Before Break (LBB) analysis discussed in Chapter 3 of- ! the AP600 SSAR (Ref. 1). This leakage limit ensures ! appropriate action can be taken before the integrity of the i lines is impaired. ' LBB is an argument which allows elimination of design for i dynamic load effects of >ostulated pipe breaks. The t fundamental premise of L38 is that the materials used in i nuclear plant piaing are strong enough that even a large ; throughwall crac( leaking well in excess of rates detectable by present leak detection systems would remain stable, and would not result in a double ended guillotine break under maximum loading conditions. The benefit of LBB is the ; elimination of pipe whip restraints, jet impingement effects, subcompartment pressurization, and internal system blowdown loads. J; As described in Section 3.6 of the AP600 SSAR (Ref.1), LBB' has been applied to the main steam line pipe runs inside containment. Hence, the potential safety significance of secondary side leaks inside containment requires detection i and monitoring of leakage inside containment. This LC0 protects the main steam lines and main feedwater lines inside containment against degradation, and helps assure that serious leaks will not develop. The consequences of violating this LCO include the possibility of further degradation of the main steam lines, which may lead to pipe break. APPLICABLE The safety-significance of plant leakage inside containment SAFETY ANALYSES varies depending on its source, rate, and duration. Therefore, detection and monitoring of plant leakage inside containment are necessary. This is accom)lished via the instrumentation required by LCO 3.4.10. "(CS Leakage , Detection' Instrumentation," and the RCS water inventory ! balance (SR 3.4.8.1). Subtracting RCS leakage as well as j (continued) !
. _ - =
I i b AP600 mA3340su aupt:16480797 B 3.7 34 4
. . . _ . - _ _ . . _ . _ , _. _ . . _ _ , m-
Main Steam Line Leakage l B 3.7.8 BASES APPLICABLE any other identified non RCS leakage into the containment SAFETY ANALYSES area from the total plant leakage inside containment (continued) provides qualitative information to the operators regarding possible main steam line leakage. This allows the operators to take corrective action should leakage occur which is detrimental to the safety of the facility and/or the public. Although the main steam line leakage limit is not required by the NRC Policy Statement criteria, this specification has been included in Technical Specifications in accordance with NRC direction (Ref. 2). LCO Main steam line leakage is defined as leakage inside i containment in any portion of the two (2) 28" I.D. main steam line pipe walls. Up to 0.5 gpm of leakage is allowable because it is below the leak rate for LBB analyzed cases of a main steam line crack twice as long as a crack leaking at ten (10) times the detectable leak rate under normal operating load conditions. Violation of this LCO i could result in continued degradation of the main steam l line. APPLICABILITY Because of elevated main steam system temperatures and pressures, the mtential for main steam line leakage is greatest in MODES 1, 2, 3, and 4. In MODES 5 and 6, a main steam line leakage limit is not provided because the main steam system pressure is far lower, resulting in lower stresses and a reduced potential for leakage. In addition, the steam generators are not the primary method of RCS heat removal in MODES 5 and 6. (continued) AP600 mA3349w wept.lb 040897 8 3.7 35 d
Main Steam Lina Leakage B 3.7.8 l BASES (continued) l ACTIONS A.1 and A.2 With main steam line leakage in excess of the LC0 limit, the unit must be brought to lower pressure conditions to reduce i the severity of the leakage and its potential consequences. The reactor must be placed in MODE 3 with 8 hours and MODE 5 within 48 hours. This action reduces the main steam line pressure and leakage, and also reduces the factors which tend to degrade the main steam lines. The Com)letion Time
@ of 4 hours to reach MODE 3 from full power wit 1out challenging plant systems is reasonable based oLg)erating experience. Similarly, the Completion Time of f 3 lours to reach MODE 5 without challenging plant systems (is also reasonable based on operating experience. In MODE 5, the pressure stresses acting on the main steam line are much lower, and further deterioration of the main steam line is less likely.
SURVEILLANCE SR 3.7.8.1 REQUIREMENTS Verifying that main steam line leakage is within the LCO limit assures the integrity of those lines inside containment is maintained. An early warning of main steam line leakage is provided by the automatic system which monitar the cintainment sump level. Main steam line leakage would appear as unidentified leakage inside containment via this system, an.d can only be positively identified by inspection. However, by performance of an RCS water inventary balance (SR 3.4.8.1) and evaluation of the cooling and chilled water systems inside containment, determination of whether the main steam line is a potential source of unidentified leakage inside containment is possible. REFERENCES 1. AP600 SSAR, Section 3.6.
- 2. NRC letter Diane T. Jackson to Westinghouse (Nicholas J.
Liparulo), dated September 5,1996, " Staff Update to Draft Safety Evaluation Report (DSER) Open Items (01s) l Regarding the Westinghouse AP600 Advanced Reactor
- Design," Open Item #365.
t h _ ,AP6,0,0_
., ,,, B 3.7 36
o** l l l FAX to DINO SCALETTI l April 7,1997 l i CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay ) Don Hutchings ! Bob Vijuk Brian McIntyre OPEN ITEM #267 (M9.4.2-3) 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 juut 54 calendar days away (40 business days). The relevant documentation related to Open Item #267 (M9.4.2-3) is SSAR Subsection 9.4.2.2.1 (pertinent pages are attached). This material was submitted to you in Revision 7 of the SSAR on May 6,1996 (more than 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." Jim Winters 412-374-5290 1 I l 1 l l l i c b
AP600 Open Items Tracking System D;tabase: Execztiv2 Sunnasary Date: 4/9/97
- Selecties: [ item no] between 267 And 267 Sotted by item #
Item DSER Section Titic/ Description Resp NRC (W) No Branch Detail Status Engineer Status Status
- Question .. .
Type. . - _ . . - . - . _ . - .-_ . .._ _ Le
- .t_te_.r N_o. _I .. - .._ Date_
267 NRR/SPLB 942 MTG-OI Winters /BRC Closed Action W
~
lM9.4.2-3 (ANNEX / AUX. BUILDINGS NON RADK5A' CTIVE IIVAfSYSTEMS} W5MDrawing VXS M0037Re$idcd[E Section ~~ 9.2.2.4 and Tables 9.41-6 and 9.4.2-7 of the SSAR show two AllUs per LIIVAC subsynem, while Figwe 9.42-3 of the SSAR shows only one AlfU Clarify the number of AllU provided in each LIIVAC subsystem and revise affected docenents and figures accordingly. _
~ ~
' Closed - SSIllimb'section 9 4.2XRevision 7. stales that fiw each subsysicm in the annex / ausi5iktdiIdks nonradioactive llVAC specm, then_ _ _ . _
f are two air handling units. They are described with each subsystem. The applicable figures and tables have been made consissent w d i '5i P'i "5-D L._. V i i b F I l Pege: 1 Total Records: I ;
- 9. Auxiliary Systems Non-safety electrical penetration rooms (auxiliary building) .... ... . 50-105 Reactor trip SWGR rooms (auxiliary building) . . . ........ .. . . 50-105 Valve / piping penetration room (auxiliary building) . . . .. . . 50-105 l Ancillary diesel generator room (annex building) .. .. . . 50-105 Upset Conditions (Loss of Plant ac Electrical System)
Switchgear rooms (annex building) . . . . .. .. ..... 122 (maximum) l Battery charger rooms (annex building) . . .. .. .. 122 (maximum) i Ancillary diesel generator room (annex building - DG sets operating) 122 (maximum) 9.4.2.2 System Description The annex / auxiliary buildings nonradioactive HVAC system consists of the following independent subsystems: '
- General area HVAC' subsystem Switchgear room HVAC subsystem Equipment room HVAC subsystem MSIV compartment HVAC subsystem Mechanical equipment areas HVAC subsystern e Valve / Piping penetration room HVAC subsyst.m The defense in dep*lt portie of the system is shown in Figure 9.4.2-1.
9.4.2.2.1 General Description 9.4.2.2.1.1 General Area HVAC Snh7'- area HVAC subsystem serves personnel areas in the annex building outside the security area. The general area HVAC subsystem consists of two 50 percent capacity supply I air handling units of about 5,100 scfm each, a humidifier, a ducted supply and retum air system, diffusers and registers, exhaust fan, automatic controls, and accessories. The air units are located on the Sw roof of the anaex building at elevation 117'-6" units dischar *traeagtm; led supply distribution rystem which is routed thm ;'. m outiding i to provide air inte, th . various rooms ano anc xivcu via registers. An electric heating coil
)
is provided in the b=nch suppl.y duct to the men's and women's change rooms for tempering the supply air. A humidifier is provided in the system to provide a minimum space relative humidity of 35 percent. Air from the rr . and women's locker, toilet, and shower facilities in the annex building is exhausted direct y to atmosphere by an exhaust fan. Room air from the remaining areas served is recirculated back to the air handling unit via a ceiling return plenum and a retum I h Revision: 11 3 WSStkighouse 9.4-17 February 28,1997
!;8i:
- 9. Auxiliary Systems l
duct system. Outside make-up air is added to the return air stream at the air handling units O to replace air exhausted from toilets and showers in the area served. 9.4.2.2.1.2 Switchgear Room HVAC Subsystem
~
The switchgear room HVAC subsystem serves electrical switchgear rooms in the annex 4 building. He switchgear room HVAC system' consists of two 100 percent capacity air handling units, a ducted supply and retum air system, and automatic controls and accessories.
- Mr handling units are located in the north air handling equipment room in the anner 15uilding at elevauon 13Y-3". The air handlmg units discharge into a common duct distribution system that is routed through the building to the rooms served. Air is returned to the air handling units from the rooms served by a retum duct system.
9.4.2.2.1.3 Equipment Room HVAC Subsystem equipment room HVAC subsystem serves electrical and mechanical equipmen in the annex and auxiliary buildings. His subsystem also serves the security area offices , the central alarm station in the annex building. ne equipment room HVAC system consists of two 100 percent capacity air handling units, two battery room exhaust fans, a toilet exhaust fan, a duc'ted supply and return air system, and automatic controls and accessories. n The air handling units are located in tne norm air hanaimg equwumm room m me annex building at elevation 135'-3". De air handling units discharge into a common duct (l distribution system that is routed through the buildings to the various areas served. Air is retumed to the air handling units from the rooms served (except the battery rooms and rest rooms) by a retum duct system. Electric reheat coils are provided in the ductwork to areas requiring close temperature control such as the security area offices and the central alarm station. Hot water unit heaters are provided in the north air handling equipment room to maintain the area above 50*F. A humidifier is provided in the branch duct to the security areas to provide a minimum space relative humidity of 35 percent. Each non-Class IE battery room is provided with an individual exhaust system to prevent the buildup of hydrogen gas in the room. Each exhaust system consists of an exhaust fan, an exhaust air duct and gravity back draft damper located in the fan discharge. Air supplied to the battery rooms by the air handling units is exhausted to atmosphere. Air from the rest rooms is exhausted to atmosphere by a separate exhaust fan. 9.4.2.2.1.4 MSIV Compartment HVAC Subsystem De main steam isolation valve compartment HVAC subsystem serves the two main steam isolation valve compartments in the auxiliary building that contain the main steam and feedwater lines routed between the containment and the turbine building. Each compartment is provided with separate heating and cooling equipment. Revision: 11 3 h February 28,1997 ' 9.4-18 3 7le dugliOUS8
~
j 9. Auxiliary Systems l O duct system. Outside make-up air is added to the retum air stream at the air handling units ~ to replace air exhausted from toilets and showers in the area served. 9.4.2.2.1.2 'tchgear Room HVAC Subsystem l The switchgear room HVAC subsystem serves electrical switchgear rooms in the anne l building. The switchgear room HVAC system consists of two 100 percent capacity air, ; Mndline units, a ducted supply and retum air system, and automatic controls and accessories. x The air handling units are located in the north air handling equipment room in-the-annex building at elevation 135'-3". The air handling units discharge into a common duct distribution system that is routed through the building to the rooms served. Air is retumed to the air handling units from the rooms served by a return duct system. 9.4.2.2.1.3 Equipment Room HVAC Subsystem ne equipment room HVAC subsystem serves electrical and mechanical equipment rooms in
, the annex and auxiliary buildings. This subsystem also serves the security area offices and the central alarm station in the annex building. The equipment room HVAC system consists of two 100 percent capacity air handling units, two battery room exhaust fans, a toilet e hn. a duc~ted supply and retum air system, and automatic controls and accessorie s
ne air handling units are located in the north air handling equipment room in the annex 3 building at elevation 135'-3". The air handling units discharge into a common duct ' distribution system that is routed through the buildings to the various areas served. Air is retumed to the air handling units from the rooms served (except the battery rooms and rest rooms) by a retum duct system. Electric reheat coils are provided in the ductwork to areas requiring close temperature control such as the security area offices and the central alarm station. Hot water unit heaters are provided in the north air handling equipraent room to maintain the area above 50'F.
)
A humidifier is provided in the branch duct to the security areas to provide a minimum space relative humidity of 35 percent. Each non-Class IE battery room is provided with an individual exhaust system to prevent the buildup of hydmgen gas in the room. Each exhaust system consists of an exhaust fan, an exhaust air duct and gravity back draft damper located in the fan discharge. Air supplied to the battery rooms by the air handling units is exhausted to atmosphere. Air from the rest rooms is exhausted to atmosphere by a separate exhaust fan. 9.4.2.2.1.4 MSIV Compartment HVAC Subsystem The main steam isolation valve compartment HVAC subsystem serves the two main steam isolation valve compartments in the auxiliary building that contain the main steam and feedwater lines routed between the containment and the turbine building. Each compartment is provided with separate heating and cooling equipment. Revision: 11 ' b 3 February 28,1997 9.4-18 T Westinghouse
' .y ..
- 9. Auxiliary Systems The main steam isolation valve compartment HVAC subsystem consists of two 100 percent I
capacity supply air handling units of abou' 3,300 scfm each with only low efficiency filters, ducted supply air distribution, automatic controls, and accessories for each m i
'% ion valve compartment.
/
Tne air handling units are located directly within the space served. One unit in each compartment normally operates to maint_in the temperature cf the comparttry.nt. The air handling units can be connected to the standby power system, for investment protection, in the event of loss of the plant ac electrical system. 9.4.2.2.1.5 Mechanical Equipment Areas HVAC Subsystem I The mechanical equipment areas HVAC subsystem serves the ancillary diesel generator room, demineralized water decxygenating room, boric acid batching / transfer rooms, and air handling equipmen1 rooms in the south end of the annn hnilmng --
~ __
I ( ne mechanical equipment areas HVAC subsystem consists of two 50 percent capacity air i handling units with supply fans and retum/ exhaust fans of about 2,200 scfm each, a ducted supply and return air system, automatic controls, and accessories. The air handling units are located in the lower south air handling unit equipment room on elevation 135'-3" of the annex building. l The ancillary diesel generator room is supplied air from the xt handling units to maintain I normal design temperatures. Air supplied to the room is exhausted direct to outdoors by I means of a separate exhaun fan. Ventilation and cooling for the room when the ancillary I diesel generators operate is provided by means of manually operated dampers and opening i doors to allow radiator discharge air to be exhausted direct to outdoors. 9.4.2.2.1.6 Valve / Piping Penetration Room HVAC System 7 tration room HVAC subsystem serves the valve / piping penetration room j on elevation 100'-0" of the auxiliary building. He valve / piping penetration room HVAC l subsystem consists of two 100 percent capacity air handling units wii supply fans of 1,800 i scfm each, a return air duct system, automatic controls and accessories.
~ ~
f
'Ihe air handhng units are located directly within the space served.
9.4.2.2.2 Component Description The annex / auxiliary buildings HVAC system is comprised of the following major components. These components are located in buildings on the Seismic Category I Nuclear Island or in the annex building. He seismic design classification, safety classification and principal construction code for Class A, B, C, or D components are listed in Section 3.2. Tables 9.4.2-1 and 9.4.2 2 provide the design parameters for major defense.in-depth components of the system.
)lo Revision: 11 3 W95tktgh00S8 9.4-19 February 28,1997
1 J
** TX CONFIRMATION REPORT ** AS OF APR 9 '97 15:32 PAGE.01 APG00 DESIGN CERT l
l DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 4/ 9 15:28 301 504 2300 G3--S 03'47 06 OK I I
r l' l, FAX to DINO SCALETTI 1 April 7,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay Don liutchings Bob Vijuk Brian McIntyre OPEN ITEM #265 (M9.4.2-1) 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 54 calendar days away (40 business days). The relevant documentation related to Open Item #265 (M9.4.2-1) is SSAR Subsection 9.4.2.1.2 (pertinent pages are attached). This r.aterial was submitted to you in Revision 7 of the SSAR on May 6,1996 (more than i1 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 l l . 4 l +
# 4 AP600 Open Items Tracking Systems Dabbese: Execctiv2 Samanary pat;: 4/9/97 _
Selection: I {ttem no] between 265 And 265 Sorted by item # [ leem DSER Section Titic/Descnytion . Rc5P (%)- NRC No. Branch Question Type . Deta! Status . Engineer Status ' Status . 12tter No. / . Date 215 NRR/SPLB 94.2 MTG4)I Winters /BRC Closed Action W {M9.4.2-1 (ANNEX / AUX. BUILDINGS NON-RADIOACTIVE IIVAC SYSTEMS) What is the pmtection limit for the buildup of hydrogen
! concentration in non< lass IE banery rooms in the annex 1 budding' Provide the ambient summer and winter design temperatures for which the
[VXS subsystems are designed._ _ _ _ _ . i _ - -__ ~
~ _~ "~ ~
'closedTSSARsubddiAf4.2lifRevisid7[includesthepdtectir 'imTfor buikof h[M and__ '
'referenas so the aminent summer and winter design temperar**t ue !w dessgn.,
i l N ' h i
\ t i
f i t I
't I
i I i i 4 l' age: 1 Total Records: I
o 1
- 9. Auxiliary Systems
)
1 9.4.2 Annex / Auxiliary Buildings Nonradioactive HVAC System The annex / auxiliary buildings nonradioactive HVAC system serves the nonradioactive ; I personnel and equipment areas, electrical equipment rooms, clean corridors, the ancillary ; I diesel generator room and demineralized water deoxygenadng room in the annex building, and the main steam isolation valve compartments, reactor trip switchgear rooms, and piping and electrical penetration areas in the auxiliary building. 9.4.2.1 Design Basis 9.4.2.1.1 Safety Design Basis The annex / auxiliary buildings nonradioactive HVAC system serves no safety-related function and therefore has no nuclear safety design basis. 9.4.2.1.2 Power Generation Design Basis
'Ihe annex / auxiliary buildings nonradioactive HVAC system provides the following specific functions:
Provides conditioned air to maintain acceptable temperatures for equipment and i personnel working in the area
~
Provides suitable environmental conditions for equipment in the main steam isolation valve (MSIV) compartments ('l-
=
Prevents the buildup of hydrogen in non-Class IE battery rooms to less than 2 percent hydrogen by volume
= Removes vitiated air from locker, toilet, and shower facilities 1
ystem maintains the following room temperatures based on maximum and mini normal outdoor air temperature conditions shown in Chapter 2, Table 2-1: Room or Area Temperatures ('F) Normal Operation i Offices, corridors (annex building) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73-78 Locker rooms, toilet rooms (annex building) . . . . . . . . . . . . . . . . . . . . . . . . 73-78 Central alarm station, security access area (annex building) .............7378 Non-Class 1E battery rooms (annex building) . . . . . . . . . . . . . . . . . . . . . 60-90 Switchgear and battery charger rooms (annex building) . . . . . . . . . . . . . . . 50-105 HVAC and mechanical equipment rooms (annex building) . . . . . . . . . . . 50-105 MSIV compartments (auxiliary building) . . . . . . . . . . . . . . . . . . . . . . .. 50-105 . (
/
Revision: 11 p< February 28,1997 M' v.4-16 3 W8Stingh00S8
- 9. Auxiliary Systems doet system. Outside make-up air is added to the retum air stream at the air handling units O
to replace air exhausted from toilets and showers in the area served. 9.4.2.2.1.2 Switchgear Room HVAC Subsystem The switchgear room HVAC subsystem serves electrical switchgear rooms in the annex building. He switchgear room HVAC system consists of two 100 percent capacity air handling units, a ducted supply and retum air system, and automatic controls and accessories. The air handling units are located in the north air handling equipment room in the annex building at elevation 135'-3". The air handling units discharge into a common duct distdbution system that is routed through the building to the rooms served. Air is returned to the air handling units from the rooms served by a return duct system. 9.4.2.2.1.3 Equipment Room HVAC Subsystem The equipment room HVAC subsystem serves electrical and mechanical equipment rooms in the annex and auxiliary buildings. This subsystem also serves the security area offices and the central alarm station in the annex building. He equipment room HVAC system consists of two 100 percent capacity air handling units, two battery room exhaust fans, a toilet exhaust fan, a ducted supply and return air system, and automatic controls and accessories. The air handling units are located in the north air handling equipment room in the annex building at elevation 135'-3". He air handling units discharge into a common duct h distribution system that is routed through the buildings to the various areas served. Air is ! retumed to the air handling units from the rooms served (except the battery rooms and rest ! rooms) by a retum duct system. Electric reheat coils are provided in the ductwork to areas ' requiring close temperature control such as the security area offices and the central alarm station. Hot water unit heaters are provided in the north air handling equipment room to maintain the area above 50*F. i A humidifier is provided in the branch duct to the security areas to provide a minimum space relative humidity of 35 percent. l 1 Each non-Class IE battery room is provided with an individual exhaust system to prevent the buildup of hydrogen gas in the room. Each exhaust system consists of an exhaust fan, an exhaust air duct and gravity back draft damper located in the fan discharge. Air supplied to the battery rooms by the air handling units is exhausted to atmosphere. Air from the rest _ rooms is exhausted to atmosphere by a separate exhaust fan. 9.4.2.2.1.4 MSIV Compartment HVAC Subsystem He main steam isolation valve compartment HVAC subsystem serves the two main stcam isolation valve compartments in the auxiliary building that contain the main steam and feedwater lines routed between the containment and the turbine building. Each companment is provided with separate heating and cooling equipment. Revision: 11 February 28,1997 Mf 9.4-18 T Westinghouse
- 2. Site Characteristics l
Table 2-1 (Sheet 1 of 2) l SITE PARAMETERS USED AS A BASIS FOR DESIGN CERTIFICATION l Air Temperature Limits 1 Maximum Safety") ll5'F dry bulb /80'F coincident wet bulb ! 81'F wet bulb (noncoincident) l Minimum Safety") -40*F Maximum Normal *) 100*F dry bulbn7'F coincident wet bulb l 80*F wet bulb (noncoincident)") Minimum Normal *) -10*F Wind Speed Limits Operating Basis 110 mph; importance factor 1.11 (safety),1.0 (nonsafety) Tornado 300 mph Seismic SSE 0.30g peak ground acceleration "' Fault Displacement Potential None N Soil i Bearing Strength Soils must support the AP600 under specified conditions. The average static bearing reaction due to the dead weight of the AP600 nuclear island is about 8000 pounds / square foot; the maximum static bearing reaction at a corner is about 12,000 pounds per square foot. Shear Wave Velocity Greater than or equal to 1000 ft/sec based on low strain best estimate soil properties Liquefaction Potential None i ( Revision: 10 <d December 20,1996 2-14 W W 6 5ieuse
4 o
** TX CONFIRMATION REPORT ** AS OF APR 9 '97 15:27 PAGE.01 APSOO DESIGN CERT DATE TIME TO/FROM MODE MIN /SEC PGS STATUS 01 4/ 9 15:24 301 504 2300 G3--S 02"53 05 OK O
i
4 1 FAX to DINO SCALETTI ) l April 7,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay Gordon Israelson Don Hutchings Bob Vijul: Brian McIntyre P OPEN ITEM #274 (M9.4.2-10) 1 To meet the SECY-97-051 schedule of " Applicant Submits Final SSAR Revisions & Documentation" by 5/97, we believe thst NRC must acknowledge receipt of all , Westinghouse submittals by May 30,1997. This is just 54 calendar days away (40 business days). The relevant documentation related to Open Item #274 (M9.4.2-10) is SSAR Subsection 9.4.2.2.1.6 (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 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 l ft
> t AP600 Open Item Tracking System Database: Executive Summary - Date: 43/97 Selecties: [ item nol between 274 And 274 Sorted in item 8 -
' licm DSER Section Resp (W) NRC Titic.% ion No Branch Question . _.- . Type .. . . - Detail - . . .Status
-.---.._-.-.--..- Engineer . Status ' . Status
. . . .- . - - - - - - - - -.- -- - .. - - -Lette r. -. o.. N . /- ..- . Dese - -
'274 ~ NRR/SPLB 9.4.2 MTG-OI Winters /BRC Closed Action W
~
[M9.4.2-10 (ANNEX /AUf BU'l_ l DINGS OslRADIOdCTiVilNIdIYSTEMS) TdIc~ 9I4.2I7 EtlEY5NR for ilEialve/pipdpediraUon^
;roosn llVAC system shows 2-100 percent AllU while Figure 9.4.2-3 shows a single AllU. Reconcile the difference and revisc the SSAR
[accordingly
~ ' ' ~
." Closed - SSAR'suinectUm 9I412 II Rcvision 7. states that the valve /pipk peNUAtdiM INAC
! subsystem consists of two 100 percent capacity air handimg units. lhe SSAR is now consistent . - . - . . - .
N . t Le < i i , i i Page: 1 Total Records: I
.4 0
- 9. Auxiliary Systems ne main steam isolation valve compartment HVAC subsystem consists of two 100 percent I capacity supply air handling units of about 3,300 scfm each with only low efficiency filters, ducted supply air distribution, automatic controls, and accessories for each main steam isolation valve compartment.
He air handling units are located directly within the space served. One unit in each compartment normally operates to maintain the temperature of the compartrneat. The air handling units can be connected to the standby power system, for mvestment protection, in tia event of loss of the plant ac electrical system. 9.4.2.2.1.5 Mechanical Equipment Areas HVAC Subsystem I ne mechanical equipment areas HVAC subsystem serves the ancillary diesel generator room, demineralized water deoxygenating room, boric acid batching / transfer rooms, and air handling equipment rooms in the south end of the annex building. The mechanical equipmen't areas HVAC subsystem consists of two 50 percent capacity air I handling units with supply fans and retum/ exhaust fans of about 2,200 scfm each, a ducted supply and return air system, automatic controls, and accessories. He air h'andling units are located in the lower south air handling unit equipment room on elevation 135'-3" of the annex building. I ne ancillary diesel generator room is supplied air from the air handling units to maintain I normal design temperatures. Air supplied to the room is exhausted direct to outdoors by I means of a separate exhaust fan. Ventilation and cooling for the room when the ancillary I diesel generators operate is provided by means of manually operated dampers and opening I annn - "r r f!: -- f!M G: "- m M exhausted direct to outdoors. 9.4.2.2.1.6 Valve / Piping Penetration Room HVAC System he valve / piping penetration room HVAC subsystem serves the valve / piping penetration room , on elevation 100'-0" of the auxiliary building. He valve / piping penetration room HVAC I subsystem consists of two 100 percent capacity air handling units with supply fans of 1,800 I scfm each, a retum air duct system, automatic controls and accessories. The air handimg units are located directly within the space served. 9.4.2.2.2 Component Description ne annex / auxiliary buildings HVAC system is comprised of the following nujor components. These components are located in buildings on the Seismic Category I Nuclear Island or in the annex building. He seismic design classification, safety classification and principal construction code for Class A, B, C, or D components are listed in Section 3.2. Tables 9.4.2-1 and 9.4.2-2 provide the design parameters for major defense-in-depth components of the system. l Revision: 11 February 28. 397 T Westkighouse h3 9.4-19 1
C 0,
** TX CONFIRMATION REPORT ** AS OF APR 9 '97 15:23 PAGE.01 APG00 DESIGN CERT DATE TIME TO/FROM MODE MIN /SEC PGS STATUS 01 4/ 9 15:21 301 504 2300 G3--S 01'44 03 OK l
1 I l
h I FAX to DINO SCALETTI April 7,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay Gordon Israelson Don Hutchings Bob Vijuk Brian McIntyre OPEN ITEM #297 (M9.4.8-6) 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 54 calendar days away (40 business days). The relevant documentation related to Open Item #297 (M9.4.8-6) is SSAR Subsection 9.4.8.1 (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 provide dermitive 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 I l
a; V .. AP600 Open It;m Tracking Systne Database: Executiva Summary . Dati: 4MM7 Selecties: { item nol between 297 And 297 Sorted by item # I Item DSLR Section Resp Tide / Description (W) NRC No. Branch _ gstion -Type Detail Status Engineer Status Status Leuer No / _ Dnee ., 297 NRR/SPLB 9.4.5 MTG-O! Winters /BRC Closed Aceion W NTD-NRC-95-4464
~ - -~
r lM9 4.8-6 (RADWASTE BUILDING llVAC Si' STEM) %1ia~t areihe ambientitustmc' andiinier desigritchatures fAr~which't' h iNXS t @ D _"8N8 Y '_ _.. __ _ _ _ .__ ._ _ _.__ _ _ - _ ____ ._.. _. _ . .._
- -~ ~~
~
{ Closed [- 1his~picifiidesign room temp' erasses a' rid a ref5inc~e to the mn5[nt icmpiraItures are Inclinled iriSSAR'subsc55on , N k b 9 t Page: 1 Total Records: I _ . . _ _ _ _ _ _ - _ . . _ _ _ - _ _ _ - . = _ _ - _ _ _ - _ _ _ .- __~. . . . - - - . . - . -, ... -. .. - _ . - - . . . - . _ . . - _ _ _ _ - - - _ _J
i ' ' ' d: O. Auxiliary Systems
]
9.4.8.1 Design Basis 9.4.8.1.1 Safety Design Basis The radwaste building HVAC system serves no safety-related function and therefore has no nuclear safety design basis. 9.4.8.1.2 Power Generation Design Basis he radwaste building HVAC system provides the following functions: Provide conditioned air to work areas to maintain acceptable temperatures for equipnient and personnel working in the areas
=
Provide confidence that air movement is from clean to potentially contaminated areas to minimize the spread of airbome contaminants
=
Collect the vented discharges from potentially contaminated equipment
=
Provide for radiation monitoring of exhaust air prior to release to the environment
=
Maintain the radwaste building at a negative pressure with respect to ambient to prevent unmonitored releases from the radwaste building be system maintains the following temperature based on maximum and minimumalno outdoor air tempe.rature conditions shown below in Chapter 2, Table 21: Room or Area Temperatures j (*F) Processing ateas and storap areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50- 105 Mechanical and electrical equipment rooms ....... .... ..... . . 50-1 9.4.8.2 System Description 9.4.8.2.1 General Description
%e radwaste building HVAC system is a once-through ventilation system that consists of two integrated subsystems: the radwaste building supply air system and the radwaste building exhaust air system. The systems operate in conjunction with each other to maintain temperatures in the areas served while controlling air flow paths and building negative pressure.
I he supply air system consists of two 50 percent capacity air handling units of about 9,000 I scfm each with a ducted air distribution system, automatic controls, and accessories. He air handling units are located in an electrical / mechanical equipment room on elevation 100'-0" on the southwest side of the buildin,9 Each unit draws 100 percent outdoor air through 3M nevision: 11 February 28,1997 W Westinghouse 9.4-51
~ ~ --- : :
- 2. Site Characteristics r!
Table 2-1 (Sheet 1 of 2) I _ SITE PARAMETERS USED AS A BASIS FOR DESIGN CERTIFICATION Air Temperature Limits 1 Maximum Safety") 115'F dry bulb /80'F coincident wet bulb ' 81'F wet bulb (noncoincident) Minimum Safety") -40*F l j Maximum Normal *) 100'F dry bulbn7'F coincident wet bulb 80*F wet bulb (noncoincident)'*
-10*F Qiinimum Normal *)
Wind Speed Limits Operating Basis 110 mph; importance factor 1.11 (safety),1.0 (nonsafety) Tornado ' 300 mph Seismic SSE 0.30g peak ground acceleration "' Fault Displacement Potential None N Soil . Bearing Strength Soils must support the AP600 under specified conditions. 'Ihe average static bearing reaction due to the dead weight cf the AP600 nuclear island is about 8000 pounds / square foot; the maximum static bearing reaction at a corner is about 12,000 pounds per square foot. Shear Wave Velocity Greater than or equal to 1000 ft/sec based on low strain best estimate soil properties Liquefaction Potential None I i l 1 i I Revision: 10 h December 20,1996 2-14 3 W6smiliGUS8
J J
** TX CONFIRMATION REPORT ** AS OF APR 9 '97 15:20 PAGE,01 APG00 DESIGN CERT DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 4/ 9 15:18 301 504 2300 G3--S 02'11 04 OK
FAX to DINO SCALETTI April 8,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay Don liutchings Bob Vijuk Brian McIntyre OPEN ITEM #269 (M9.4.2-5) 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 53 calendar days away (39 business days). The relevant documentation related to Open Item #269 (M9.4.2-5) is SSAR Subsection 9.4.2.2.1.1 (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). In addition, Table 3.2-3 was revised in SdAR Revision 11 to list the classification of the toilet exhaust fans. 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 l c/W
s *l
- AP600 Open Item Tracking Syst;m Database: Executiva Summary ' Datz: 4/8/97 Selecties: [ item no] between 269 And 269 Sorted by item e leem D5ER Section Resp Titic/Desenption (W)- NRC No. Branch Questen Type Detail Status Engineer . Status Status i.etter No. / Dane 269 NRR/SPLB 9.42 MTG4)I Winters /BRC Closed Action W
~ - ~
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,M9 4.2-5 (ANNEX /AUf BUILDINUS NON-li DIOACflVIfiSAUS STEidS) Prodde nEn's and women's locker'rneihaust fans' data for~7 '
[the general area IIVAC system in Table 9.4.2-2 of the SSAR.
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[ Closed!55AUs uEionUU.l.1, Revision 7, includes a desdigEEo'f50pdion of the toilet cihmist' fans? lIhese fus are not in a safety-related or defence-inalepth portion of the system. They are not included in the 1 equipment tables or system sketches provided for the system. N D . t l t i Page: 1 Total Records: I i
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Non-safety electrical penetration rooms (auxiliary building) . 50-105 Reactor trip SWGR rooms (auxiliary building) . . . . . . . . 50-105 Valve / piping penetration room (auxiliary building) .. . . . . 50-105
- 1 Ancillary diesel generator room (annex building) . . ... .
50-105 I Upset Conditions (Loss of Plant ac Electrical System) Switchgear rooms (annex building) .. .... ... . .. 122 (maximum) Battery charger rooms (annex building) . . ........ .. 122 (maximum) 1 I Ancillary diesel generator room (annex building - DG sets operating) 122 (maximum) ; 9.4.2.2 System Description l The annex / auxiliary buildings nenradioactive HVAC system consists of the following independent subsystems:
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General area HVAC subsystem
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Switchgear room HVAC subsystem
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Equipment room HVAC subsystem MSIV compartment HVAC subsystem j Mechanical equipment areas HVAC subsystem ; Valve / Piping penetration room HVAC subsystem j l The defense in depth portion of the system is shown in Figure 9.4.2-1. i 1 9.4.2.2.1 General Description 9.4.2.2.1.1 General Area HVAC Subsystem I The general area HVAC subsystem serves personnel areas in the annex building outside the security area. The general area HVAC subsystem consists of two 50 percent capacity supply I air handling units of about 5,100 scfm each, a humidifier, a ducted supply and return air system diffusers and registers, exhaust fan, automatic controls, and accessories. The air handling units are located on the low roof of the annex building at elevation 117'-6". 'Ihe units discharge into a ducted supply distribution system which is routed through the building I to provide air into the various rooms and areas served via registers. An electric heating coil is provided in the branch supply duct to the men's and women's change rooms for tempering the supply air. A humidifier is provided in the system to provide a minimum space relative humidity of 35 percent. [ Air from the men's and women's locker, toilet, and shower facilities in the annex buildm exhausted directly to atmosphere by an exhaust fan. Room air from the remainiug areas l served is recirculated back to the air handling unit via a ceiling return plenum and a return 3 Westhghouse 3)f 9.4-17 Revision: 11 February 28,1997
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I duct system. Outside make-up air is added to the retum air stream at the air handling units to replace air exhausted from toilets and showers in the area served. 9.4.2.2.1.2 Switchgear Room HVAC Subsystem The switchgear room HVAC subsystem serves electrical switchgear rooms in the annex building. The switchgear room HVAC system consists of two 100 percent capacity air handling units, a ducted supply and retum air system, and automatic controls and accessories. The air handling units are located in the north air handling equipment room in the annex building at eleva. tion 135'-3". The air handling units discharge into a common duct distribution system that is routed through the building to the rooms served. Air is retumed to the air handling units from the rooms served by a retum duct system. 9.4.2.2.1.3 Equipment Room HVAC Subsystem 1 i l The equipment room HVAC subsystem serves electrical and mechanical equipment rooms in ' the annex and auxiliary buildings. This subsystem also serves the security area offices and the central alarm station in the annex building. He equipment room HVAC system consists of two 100 percent capacity air handling units, two battery room exhaust fans, a toilet exhaust i fan, a ducted supply and retum air system, and automatic controls and accessories. The air handling units are located in the north air handling equipment room in the annex l building at elevation 135'-3". The air handling units discharge into a common duct - distribution system that is routed through the buildings to the various areas served. Air is retumed to the air handling units from the rooms served (except the battery rooms and rest rooms) by a retum duct system. Electric reheat coils are provided in the ductwork to areas requiring close temperature control such as the security area offices and the central alarm station. Hot water unit heaters are provided in the north air handling equipment room to maintain the area above 50*F. A humidifier is provided in the branch duct to the security areas to provide a minimum space relative humidity of 35 percent. Each non-Class IE battery room is provided with an individual exhaust system to prevent the buildup of hydrogen gas in the room. Each exhaust system consists of an exhaust fan, an exhaust air duct and gravity back draft damper located in the fan discharge. Air supplied to the battery rooms by the air handling units is exhausted to atmosphere. Air from the rest rooms is exhausted to atmosphere by a separate exhaust fan. 9.4.2.2.1,4 MSIV Compartment HVAC Subsystem He main steam isolation valve compartment HVAC subsystem serves the two main steam isolation valve compartments in the auxiliary building that contain the main steam and feedwater lines routed between the containment and the turbine building. Each compartment is provided with separate heating and cooling equipment. Revision: 11 February 28,1997
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- 3. Design of Structures, C:mponents, Eqdpmelt, and Systems I
Table 3.2-3 (Sheet 61 of 64) AP600 CLASSIFICATION OF MECHANICAL AND FLUID SYSTEMS, COMPONENTS, AND EQUIPhENT Tag Number Description AP600 Seismic Principal Con- Comments Class Category struction Code Central Chilled Water System (Continued) VWS-PL-V082 Fan Coolers Return B I ASME m-2 Containment Isolation VWS-PL-V086 Fan Coolers Return B I ASME m-2 Containment Isolation VWS PL-V424 Containment renetration Test B I ASME m-2 Connection VWS-PL-V425 Containment Penetration Test B I AShE W-2 Connection I or syneggenn - Or: E Annex / Auxiliary Nonradioactive Ventilation System (VXS)Imation: Auxiliary Building end Annex Building I n/a Air Handling Unit Fans Note 2 NS AMCA Providing AP600 Equipment Class D Function i n/a Dampers Providing VXS Note 2 NS ANSI /AMCA-l AP600 Equipment Class D 500 Function I n/a Fire Dampers Note 3 NS UL-555 Balance of system components are Class E or Class L l ter Heating System WYG Location: Various l System components are Class E I Diesel Generator Building Ventilation System (VZS) Location: Diesel Generator Building n/a Unit Heaters Providing Note 2 NS AMCA AP600 Equipment Clau D l Function n/a Fans Providing AP600 Note 2 NS AMCA Equipment Class D Function 1 y Westinghouse
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3.2-79 Revision: 11 February 28,1997 l i
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** TX CONFIRMATION REPORT ** AS OF APR 9 '97 15:09 PAGE.01 APG00 DESIGN CERT DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 4/ 9 15:06 301 504 2300 G3--S 02"56 05 OK
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FAX to DINO SCALETTI April 8,1997 l CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay i . Don Hutchings Bob Vijak ' Brian McIntyre OPEN ITEM #271 (M9.4.2-7) 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 53 calendar days away (39 business days). The relevant documentation related to Open Item #271 (M9.4.2-7) is SSAR Subsection 9.4.2.2.3 (pertinent pages are attached). This material was submitted to you in Revision 7 of the SSAR on May 6,1996 (more than 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." t 4 h ; Jim Winters 412-374-5290 I l l 1 I M vl i
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AP600 Open items Tracking System Database: Exec tivaSuunniary Date: 4/8/97 Selecties: [ item no] between 271 And 27I Sorted by Item # ! lice DSI.R Section Resp Title /Descnptkm (W) NRC No Branch Quesuon - Typc Detail Status Engineer Status . Status
.. _Le.t.t_er. .No_ / _ .. ,.. _ D.. at,e _~
271 NRR/SPLB 94.2 MTG4M Winters /BRC Closed Action W
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(M9 4 2-7 (ANNEX / AUX. BUILDINGS NON-RADIOACTlYE llV1C $55TEMS) FIgEe 5 2-2 ef the SSAR shows three'luh watcs unE hiascrs'4 lwith temperature switches serving the mechanical equipment room in the annen I building with a provision for the law water to be provided from the
;VYS. Addnionally, the mechanical equipment room IIVAC subsystem also serves the RCC and inadequate core cooling non-class IE penetration
, , rooms and reactor trip switchgear I and Il rooms in the musi!iary buildmg. Westinghouse needs to reflect the abose informahon with its associased
' details in the Section 9 4.2.2.2 of the SSAR.
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' Clos d - SSIR subseciidn 9I4T2il,' R'evisE7,'iEIAa description of'the operation of'heaie~r sin cIK50f _ ~ . . . _ _ _ _ _ . . _ . _ .~ ~ - _ .~~
an]neauxiliary buildings, nonradioactive llVAC subsystems. _ i 6 N L . M - I
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Electric Heating Coils i 1 He electric heating coils are multi-stage fin tubular type. The electric heating coils meet the ' requirements of UL 1096 (Reference 10). Electric Unit Heaters he electric unit heaters are single-stage or two-stage fin tubular type. The electric unit l heaters are UL-listed and meet the requirements of UL 1025 (Reference 26) and the National l Electric Code NFPA 70 (Reference 28).
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i Shutoff, Control, Balancing, and Backdraft Dampers i Multiblade, two-position pneumatically- or motor-operated shutoff dampers are parallel-blade type. Multiblade, contml and balancing dampers are opposed-blade type. Backdraft dampers are provided to prevent backflow through shut down fans. Air handling unit and fan shutoff , dampers are designed for maximum fan static pressure at shutoff flow. Dampers meet the l performance requirements of ANSI /AMCA 500 (Reference 14). i Fire Dampers I Fire dampers are provided at duct penetrations through fire barriers to maintain the fire resistance ratings of the barriers. The fire dampers meet the design arid installation requirements of UL 555 (Reference 15). Ductwork and Accessories Ductwork, duct supports and accessories are constructed of galvanized steel. Ductwork subject to fan :hutoff pressure.is stmeturally designed for fan shutoff pressuret.. Ductwork, supports and accessories meet the design and construction requirements of SMACNA High Pressure Duct Construction Standards (Reference 16) and SMACNA HVAC Duct Construction Standards - Metal and Flexible (Reference 17). 9.4.2.2.3 System Operation 9.4.2.2.3.1 General Area HVAC Subsystem Normal Plant Operation During normal plant operation, both supply air handlitig tini8'ahdte toiledsgwer exhaust fan operate continuously to maintain suitable temperatures in the areas serve 6 e temperature of the air supplied by each handling units is controlled by individual temperature i controls with their sensots located in the annex building main entrance. He temperature ( sensor sends a signal to a temperature controller which modulates the chilled water control valve and the face and bypass dampers across the supply air heating coil to maintain the area [' (, February 28,1997 Revision: 11 [ ',"le Juqiisase 9.4-21
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- 9. Auxiliary Systems O
within the design range. The switchover between cooling and heating modes is automatically controlled by theyyture controllm. ~~% )
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Supplemental heating is provided for the men's/ women's change room areas by an electnv l reheat coil located in the s'upply air duct to the areas served. The reheat coil operates i intermittently under the control of its temperature controller with sensor located in the women's change room, which modules the electric heating elements to maintain the space rature in the change room areas within the design range. The supply air is humidified by a common humidifier located in the doctwork downstream of the supply air handling units. A humidistat located in the main entrance of the annex building intermittently operates the humidifier to maintain a minimum space relative humidity of 35 percent in the area served.
'Ihe differential pressure drop across each supply unit filter bank is monitored, and individual alarms are actuated when any pressure drop rises to a predetermined level indicative of the need for filter replacement., To replace the filters on a supply unit, the affected supply fan is stopped and isolated from the duct system by means of isolation dampers. The toilet / shower exhaust fan is also stopped. During filter replacement, the system operates at approximately 50 percent capacity. This mode of operation will maintain a slight positive pressure in the building.
Abnormal Plant Operation
, _j The general area HVAC subsystem is not required to operate during any abnormal plant condition.
9.4.2.2.3.2 Switchgear Room HVAC Subsystem Normal Plant Operation During normal plant operation. one air handling unit operates continuously to maintain the oor tempeatures in the two switchgear rooms. De tem. - of the air supplied by the air handling unit is maintained at 62*F by a temperature controller bas ide ambient , temperature conditions. When the outdoor air temperature is below 62*F, the te ture l controller modulates the outside air, return air and exhaust air dampers of the air handling 't ; to mix return air and outside air in the proper proportion, and modulates the face and bypass j dampers of the hot water heating coils to maintain a mixed air temperature of 62'F. A ! nummum amount of outside air is always provided for ventilation requirements. When the i outdoor temperature is above 62'F, the outside air, retum air and exhaust air dampers automatically reposition for minimum outside air and the temperature controller modulates the l l c water control valves to maintain the supply air at 62*F. The switchover between l cooling eating modes is automatically controlled by the supply air temperature j controllers. ! l Revision: 11 y l February 28,1997 9.4-22 3 Westingh00S8 l
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- 9. Auxiliary Systems The differential pressure drop across each air handling unit filter bank is monitored and individual alarms are actuated when the pressure drop rises to a predetermined level indicative of the need for filter replacement. To replace the filters on an air handling unit, the unit is stopped md isolated from the duct system by means of isolation dampers. During filter replacement, the second air handling unit operates at full system capacity.
Abnormal Plant Operation
.7 ferent'of a loss of the plant ac electrical system, the air hand i 't supply and ActGn/ exhaust fans are connected to the standby power system to provide ventilati ' oling to the diesel bus switchgear. This cooling permits the switchgear to perform its defense depth functions in support of standby power system operation. In this mode of operation, the switchgear rooms are cooled utilizing once-through ventilation using outdoor air. When in the once-through ventilation mode, the switchgear rooms will be maintained at or below 122*F. Equipment in these rooms that operate following a loss of the plant ac electrical system are designed for continuous operation at this temperature. To maintain the areas above freezing, the mixing dampers will medulate to maintain a supply air temperature of 62"F for outdoor temperatures below 62'F. For outdoor temperature above 62*F, the outside air, return a , _.f ~ bne Mr amn- px :.c.d k. . x.x ^r-f h 9.4.2.23 3 Equipment Room HVAC Subsystem Normal Plant Operation During normal plant operation, one air handling unit and both battery room exhaust fans operate continuously to maintain the indoor temperatures in the equipment and security access areas served by the system.
De temperature ir supplied by~the air handTtET' emet is maintained at 6.* by a tempera ntroller based on outside ambient temperature conditions. When the ou r ai perature is below 62*F, the temperature controller modulates the outside air, return and exhaust air dampers of the air handling unit to mix rerum air and outside air in the proper proportion, and modulates the face and bypass dampers of the hot water heating ceils to maintain a mixed air temperature of 62*F. A minimum amount of outside air is always provided for ventilation requirements. When the outdoor air temperature is above 62 F, the outside air, retum air and exhaust air dampers automatically reposition for minimum outside air and the temperature controller modulates the chilled water control valves to maintain the supply air at 62'F. He switchover betwa & ag.ard H heating modes is automaticall alled by 6mppl7 air ternperature centrollers ' eat coils serving the security access areas are controlled by temperature controlle with sensors located in the areas served. The temperature sensor sends a signal to a temperature controller which modulates the electric heating elements to maintain the secudty access areas at their design temperatures. Hot water unit heaters operate intermittently to provide supplemental heating for the north air handling equipment room to maintain the area temperature above 50 F. l Revision: 11 3 W85tinghCUS8 (3 9.4-23 February 28,1997
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- 9. Auxiliary Systems A humidistat located in the security xces area intermittently operates the humidifier to maintain the security office area at a rrdaimum space relative humidity of 35 percent.
The differential pressure drop across each air handling unit filter bank is monitored, and individual alanns are actuated when the pressere drop rises to a predetermined level indicative of the need for filter replacement. To replace the filters of an air handling unit, the unit is stopped and isolated from the duct system by means of isolation dampers. Dunng filter re lacement, the second air handling unit cperates at full system capacity. A temperature controller opens the outside air intake and starts and stops the elevator mac i room exhaust fan as required to maintain room ces:gn temperature conditions. A local thermostat controls the electric unit heater. Abnormal Plant Uperanon in the event of a loss of the plant ac electrient evem the air handling unit supply and return / exhaust fans ar m me to the standby power system to prow -*Meon cooling to the de sw' and inverters. This cooling permits that equipment to perform i se in d unctions. In this mode of operation, the rooms are cooled utilizing once-throug ntilation using outdoor air. When in the once-through ventilation mode, the de switchgear and inverter areas will be maintained at or below 122'F. Equipment in those areas that operate following a loss of the plant ac electrical system are designed for continuors operation at this temperature. To maintain the areas above freezing, the mixing dampers will modulate [ to maintain a supply air temperature of 62*F for outdoor temperatures below 62*F. For outdoor temperature above 62*F, the outside air, retum air, and exhaust air dampers are po um.M N n once-through flow.
, 9.4.2.23.4 MSIV Compartment HVAC Subsystem Normal Plant Operation os-u operation, one of the main steam isolation valve compartme dling units in each compartment operates continuously in a recirculation mode to maintin the indoor temperature in the equipment area served by the system. A temperature controller j
modulates the chilled water and hot water control valves serving the operating unit to maintain I the compartment temperature at or less than 105'F and above a minimum of 50*F. The switchover between cooling and heating modes is automatically controlled by the area temperature controller. The differential pressure crop across eacn au u- .or, mm una oank is monitored and irdNidual alarms are actuated when the pressure drop rises to a prede: ermined level indicative of a need for filter replacement. An air handling unit may be shutdown for filter rep',acement or other maintenance as required, with the other air handling unit in the same compartment operating to maintain the area temperature. Revision: 11 February 28,1997 9,4 24 3 Westkighouse
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Abnormal Plant Operation The main steam isolation valve compartment HVAC subsystem is not required to operate during abnormal plant conditions. 9.4.2.2.3.5 Mechanical Equipment Arcas HVAC Subsystem Durine normal olate,"^-
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v._ r"~>dv to maintain the KrJfTemperatures in the areas served. The temperature of the air supplie iair handling unit is controlled by individual temperature controls with their sensors located in upper south air handling equipment room. The temperature sensor sends a signal to a l temperature controller which modulates the face and bypass dampers across the rupply air , heating coil and the chilled water control valve to maintain the mechanical equip.nent areas within the design temperature range. The switchover between cooling and hea6ng modes is automatically controlled by the area temperature controller. Differential pressure drop across each air handling unit filter bank is monitored, and individual alarms are actuated when pressure drop rises to a predetermined level indicative of the need for filter replacement. During filter replacement, the system operates at approximately 50 percent capacity. I The exhaust fan for the ancillary diesel generator room operates continuously for room I ventilation. Abnormal Plant Operation The mechanical equipment areas HVAC subsystem is not required to operate during abnormal plant conditions. I When the ancillary diesel generator sets are operated, a manual damper is opened as required j
! and the outside door is opened to mdntain acceptable temperatures.
9.4.2.2.3.6 Valve / Piping Penetration Room HVAC Subsystem Normal Plant Operation C g normal plant operation, one air handling unit operates continuously in a rectre mode to maintain the indoor temperature in the room. A temperature controller modulates 3 the chilled water control valve and opens and closes the hot water control valve serving the ! l operating unit to maintain the area temperature at or less than 105'F and above a minimum of 50'F. 'Ihe switchover between cooling and heating modes is automatically controlled by 6 area temperature controller. The differential pressure drop across each air handling unit filter bank is monitored, and individual alarms are actuated when the pressure drop rises to a predeterm'med level indicative of the need for filter replacement. i
@ Revision: 11 i T@ 9.4-25 February 28,1997 i
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- 9. Auxiliary Systems Abnormal Plant Operation e
ne valve / piping penetration room HVAC subsystem is not required to operate during abnormal plant conditions. 9.4.2.3 Safety Evaluation The annex / auxiliary buildings nonradioxiive HVAC system has no safety-related function and therefore requires no nuclear safety evaluation. 9.4.2.4 Tests and Inspections The annex / auxiliary buildings nonradioactive HVAC system is designed to permit periodic inspection of system components. Each component is inspected prior to installation. Components of each system are accessible for periodic inspection during normal plant operation. A system air balance test and adjustments to design conditions are made during the plant preoperational tes't program. Air flow rates are measured and balanced in accordance with the guidelines of SMACNA HVAC Systems - Testing, Adjusting, and Balancing (R:ference 19). Instruments are calibrated during testing. Automatic controls are tested for actuation at the proper setpoints. Alarm functions are checked for operability. 9.4.2.5 Instrumentation Applications he armex/ auxiliary buildings nonradioactive HVAC system operation is controlled by the plant control system (PLS). Refer to subsection 7.1.1 for a discussion of the plant control system. Temperature controllers and thermostats maintain the proper space temperatures. Supply air temperature is controlled by either sensing local room temperature or by sensing the supply air temperature in the air handling unit discharge duct, depending on the subsystem. Unit heaters are controlled by local thermostats. Temperature indication and alarms are accessible locally via the plant control system. Temperature is indicated for each air handling unit supply air discharge duct, except for local recirculation units such as those in the main steam isolation valve compartment and valve / piping penetration room. Operational status of fans is indicated in the main control room. De fans and air handling units can be placed into operation or shutdown from the main control room or locally. Differential pressure indication is provided for each of the filters in the air handling units and an alarm for high pressure drop is provided for each air handling unit. , Airflow is indicated for the air handling unit and exhaust fan discharge ducts. Alarms are , I provided for low air flow rates in the fan discharge ducts. i b.~ Revision: 11 February 28,1997 % 9.4-26 Y WOStktgfl00S8 I 9
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** TX CONFIRMATION REPORT ** AS OF APR 9 '97 13:36 PAGE.01 APG00 DESIGN CERT 1
I DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 4/ 9 13:21 301 504 2300 G3--S 14"45 08 OK l l l I I l l l l
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