Text

Provides W Response to FSER Open Items on AP600.Summary of Encl Provided in Table 1.W Status Will Be Changed to Confirm W
	ML20198K606
Person / Time
Site:	05200003
Issue date:	01/09/1998
From:	Mcintyre B; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	NSD-NRC-98-5515, NUDOCS 9801150055
	Download: ML20198K606 (126)
	v • d • e

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j. ! \v i-. j Westinghouse Energy Systems a an Pittstugh Prinsytoma 16730 03%

Electric Corp 0 fall 00 DCP/NRCl209 NSD NRC-98 S$l$

Docket No.: 52 003 January 9,1998 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20$$$

ATiliN110N: T. R. QUAY SulljECT: AP600 RESPONSE TO FSER OPEN ITEMS

Dear Mr. Quay:

linclosme I of this letter provides the Westinghouse responses to FSER open items on the AP600. A ,

summary of the enclosed re:.ponses is provided in Table 1. Included in the table is the FSER open -

item nuinber, the associated OITS number, and the status to be designated in tl.e Westinghouse status column of OITS. Enclosure 2 provides the ret'>ot..e for FSER open item 220.128F (OITS fl6315).

Tl.c Westinghouse status will be changed to Conurm W.

'the NRC should review the enclosures and inform Westinghotae of the status to be designated in the

NRC Status" column of OITS. t Please contact me on (412) 374 4334 if you have any questions concerning this transmittal.

$4/$

lirian A. McIntyre, Manager Advanced Plant Safety and 1,1 censing jml Enclosures ec: W. C. Iluffman, NRC (Enclowre 1)

T. J. Kenyon, NRC (Enclosure 1) . . <7 I'[ i J. M. Sebrosky, NRC (Enclosures 1 & 2) . i D C. Scaletti, NRC (Enclosure 1)

N. J. Liparuto. Westinghouse (w/o Enclosures) ,

n mamm :

DCF/NRCl209 NSD NRC 98 5515 2 January 9,1998 Table i 1 Int of FSI:R Open items included in letter l>CP/NRCl209 FSI:R Open item - OITS Number Westinshouse status in OITS 220.116F (RI) 6303 Confirm W 410 403F 6430 Action N 410.413F 6431 Confirm W 410.416F 6391 Action N 410.419F 6394 Continn W 410,420F 639f Confirm W 410.421F 6196 ConGrm W 410.424F 6399 Conurm W 410.431F 6406 Confinn W 420.12HF 6496 Confirm W 420.129F 6497 Confirm W 440.753F 6411 Confirm W 440.770F 6344 Action N 440.78317 6';65 Confirm W 440.788F 6445 Confirm W 440.793r 6443 Action N 440.804F 6478 Action N 440.80812 6483 Action N 650.25F 6327 Confirm W 650.26F 6328 Confinn W l 720.444F 6182 Confirm W 720.451F 6387 Confirm W 720.452F 6388 Confirm W 720.453F 6389 Confirm W

, 1 a.

l I:nclosure I to Westinghouse j

i.etter DCP/NRCl209 !

l January 9.1998 l

i

FSER Open hem - y.

' I

. Open item 226.116F (OITS #6303) .tsponse Revision 1 -

Westinghouse was requested to provide the basis for classifying the single story portion of me Radwaste Duilding as non seismic and the high bay area of the Radwaste Building as SC-II. The staff identified the concern of the seismic design of the radwaste building as Open item 3.7.2.8 3.

In Revision 9 of SSAR Section 3.7.2.8, Westinghouse stated that the radwaste building is classified as nonseismic and is designed to the seismic requirements of the UBC Zone 2A with an importance factor of 1.25. However, Westingtouse did not make a commitment that the collapse r ~ this building l

will not impair the safety function of the NI structures.

la Revision 12 of SSAR Section 3.7.2.8, Westinghouse, based on the energy balance theory, provided the analysis procedures which are ' be used for demonstr.aing that the collapse of the radwaste building will not cause any damage of the Ni structures. Because the application of energy balance for checking potential damages of structures is consistent with the industry practice, it is acceptable to the staff - Also, in the August 11 through 15,1997 meeting, the staff reviewed the final calculation (Calculation No. 5000-82C-002) and found that the analysis procedure described in the SSAR was properly applied in the evaluation of the impact between the Ni and the radwaste building and that the impact from the radwaste building in the event of an SSE would not impair the integrity of the N1.

His is acceptable to the staff regarding the potentialinteraction between the radwaste building and the Ni structures. However, Westinghouse re.ised its commitment and stated in Revision 17 of SSAR Section 3.7.2.8.2 that the radwaste building is designed to the seismic requirements of UBC, Zane 2A with an importance factor of 1.0 Also, the design of this building is based on the assumption that the building will collapse under an SSE. However, a collapse of the radwaste building has a potential to spill or release radwaste and cause personnel exposure. Consequently, a reasonable assurance is needed to ensure that this building will not collapse at a level of earthquake less than the SSE. Such a design would be consistent with a design based on UBC Zone 3 requirements as used for the design of the turbine building by Westinghouse. The staff previously accepted these criteria for other advanced reactors. On this basis, Westinghouse's response to Open item 3.7.2.8 3 is not acceptable.

Response: (Revision 1)

The main function of the AP600 radwaste ouilding is to store solid radwaste including dry active wades t.nd solidi'ied wastes from radwaste processing until it is processed or transferred offaite. We AP600 radwaste builomg does not contain gascon radwaste and normally does not contain liquid radwaste. There are provisW '.s ;lacement of mobile processing systems in the radwaste building to

! process some of the liquid radww.u, such as wastes from the chemistry lab and wash sinks. The mobile processing systems may have sitall tanks which hold limited quantities of liquid waste. Re-tanks that store liquid wastes for processing; are located in the auxiliary building. Processing of wastes such as filters and spent demineralizer resins that may have significant activity is done within the auxiliary ouilding not b the radwaste building.

The auxilbry building is a seismic Category I structure. De seismic requir;:ments for the AP600 l

radwaste buliding are consisten' with Position IV.4. in Branch Technical Position ETSB 113 which states that Regulatory Guide 1.6 43 seismic criteria are not applicable for structures housing mobile systems, YMN 220.116-1

i

?tER Open hem ..

De criteria used for radwaste buildings in other advance reactor designs is not applicaple for the AP600. Unlike the AP600, the radwaste buildings in those designs contain tanks and equipment used to store, process, and solidify liquid wastes.

I The change in imponance factor from 1.25 to 1.0 effective in Revision 17 was made in conformance

1. with a change in the ALWR Utility Requirements Document. The change to the Utility Requiremer.ts i Document was included in a revision after the revision reviewed by the liRC. Th safety evaluation i report for the ALWR Utility Requirements Document for Passive Plants acepts design to the seismic l requirer.nents of the UBC Zone 2A with an impottance factor of 1.25 for tL f a '. waste bultding.

I Although building collapse has little potential to spill or release radwaste for the AP600 arrangement,-

I the AP600 SSAR will be revised to be consistent with that requirement. D.s provides a reasonable I assurance that the building would not collapse at a level of canhquake belt.w the SSE, Since the staff I accepted this criteria in the Safety Evaluation Repon for the ALWR Utility Requirements Document I for Passive Plants, the criteria used for other advanced reactors does not need to be considered.

' SSAR Revision: NONE I Revise the first paragraph of subsection '3.7.2.8.2 as follows:

ne radwaste building is classified as nonseismic and is designed to the seismic requirements I of the Uniform Building Code, Zone 2A with an imponance Factor of 1.25. 0-As shown in the radwaste building general arrangement in Figure 1.2 22, it is a small steel framed building.

If it were to impact the nuclear island or collapse in the safe shutdown canhquake, it would not impair the integrity of the reinforced concrete nuclear island. He minimum clearance between the structural elements of the radwaste building above grade and the nuclear island is 4 inches.

YO 220.116-2

. FSER Open hem _..- l 410Af0F (OITS #6430) ;

Standard Review Plan (SRP) section 3.6.1 specifies that the plant design for protection against '

postu!ated pipe failures outside antainment include high and moderate-energy fluid system piping.

With regard to moderate energy piping, the staff has implemented the guidance of section 3.6.1 to emphasize that protection against the pipe failure of non seismic mod: rate unergy piping should be analyzed for both pipe breaks (e g., double ended guillotine break )ist pipe cracks.. Containment environmental consequences from such a moderate-energy pipe break could adversely impact essential equipment, which likely have not been considered in other analyses.

AP600 analysis for moderate-energy piping outside containment needs to properly consider protection against postulated pipe failures resulting from pipe breaks (De staff acknowledges that SRP section 3.6.2. " Location and Dynamic Effects of Postulated Piping Failures," provides guidance on the analysis of both seismic and non seismic moderate-energy piping for pipe cracks'only.)

Response

Westinghouse understands that the change in the guidance for evaluation of protection for moderate-energy is to consider the effect of breaks in non seismically analyzed pipes. He postulated failure in roeismically analyzed piping remains as consideration of cracks consistent with the guidance in Standard Review Plans 3.6.1 and 3.6.2.

Safety related systems required for safe shutdown are not expected to be adversely affected by the dynamic effects of postulated pipe breaks in non-seismic, moderate-energy piping. By design, non-wismic piping is not routed near safety related piping or equipment. If there is B31.1 piping whose continued function is not required, but whose failure or interaction could degrade the functioning of a safety class component to an unacceptable level, then this B31.1 piping is analyzed and designed for the SSE using the same methods as specified for seismic Category 1 piping. For example, nonsafety-related piping connected to safety related components is analyzed and designed for seismic loadings, because the piping model includes piping adjacent to the containment penetration area up to the first anchor, De effect of moderate-energy line breaks on safety-related equipment inside containment and in compartments outside of containment that include high-energy lines are bounded by the effects of the high energy lines. De compartments outside of containment that include safety related comp onents.

- moderate energy lines, and do not contain high energy lines are limited to a few rooms containing containment isolation valve in the auxiliary building and the PCS valve room located above the containment near the shield building roof. De PCS valve room does not include non wismically analyzed, moderate energy piping. He moderate-energy lines connected to the containment isolation valves are analyzed seismically from the penetration up to the anchor, ne turbine building, annex building, and the radwaste building do not contain safety related systems or components and are not evaluated.

De fire protection system piping is routed through corridors and stairwells that do not contain safety related equipment and is not evaluated for postulated breaks. Lines at or near atmospheric pressure such as drain lines or HVAC ducts are not evaluated for postulated breaks.

YNW 410.403-1

I i

PSER Open item- ,, _ . ,

I

^

The environmental effects of a moderate energy break or cract are a result of the spray of water on

. equipment. Temperature and high humidity effects are not cons!derted. The environmental effects of a -

break or crack in a compartment that_ includes a postulated high-energy rupture is bounded by the

- environmental effects of the high energy pipe rupture. See the response to FSER open item 410.413

for a discussion of moderate-energy pipe rupture environmental effects.

- SSAR Revision: NONE

(

YO 410.403 2

FSER Open item , , -

i 410.413F (OITS #6431)

Design basis events are routinely evaluated and addressed by plant design and procedures. When the design basis threat has been addressed, less-than-design basis events may then become the worst-case events and their impact should be evaluated to ensure that the consequences are bounded by the design-basis event analysis. These type of events may result in conditions not readily seen, consid-cred, or properly addressed, in such cases, it may be prudent to revisit previous assumptions about the adequacy of plant design and procedures.

With regard to protection against Duid system pipe failures for high and moderate-energy piping, AP600 analysis needs to properly consider the above type of event. (Maine Yankee Event Report 97-009-01 provides information concerning an operating plant occurrence that illustrates the staff's concern.)

Response

The event referred to above involved the potential for a crack in a moderate energy line and resultant environmental conditions in an area in the turbine building that had not been considered for environmental qualification because there were no high energy line breaks in the area. This situation is not applicable to the AP600. The turbine ouilding, annex building, and the radwaste building do not contain safety-related systems or components.

The safe shutdown systems and components required to mitigate a pipe break in the AP600 are located on the nuclear island. ihese systems and components are located in containment and to a limited extent in the auxiliary building. The safe shutdown equipment inside containment is environmentally qualified for the effects of a design basis accident. Table 3.6-2 identifies the compartments and high energy breaks that are postulated as the source of hot. high pressure Guid. His table also identifies compartments outside containment that have hot, high energy line breaks. De environmental effects of high energy line breaks on safe shutdown systems and components in these areas bound the effects of moderate energy line breaks.

Dere are limited areas in the auxiliary building with safety related equipment in the vicinity of only moderate energy lines. The safety related components to be evaluated are isolation valves and penetrations in compartments where the containment isolation valves for moderate energy systems are located and the valves in the PCS valve room. Cracks are postulated in safety-related and nonsafety.

related moderate energy lines. Temperature and high humidity effects do not have to be considered for a moderate energy break. The environmental effects of a moderate energy break or crack are a result of the spray of water on equipment. The potential for spray effects is evaluated as part of the flooding analysis. The fine protection system piping is routed through corridors and stairwells that do not contain safety related equipment and is not a postulated source of environmental effects. The effects of spray from fire fighting efforts is considered as part of the fire evaluation.

Safety-related, active valves subject to a water spray due to a crack or break in a moderate energy line are qualified to operate in the presence of the spray. He valves qualified for the environmental effects of high-energy line breaks do not require a s;parate qualification for the effect of water sprays.

Dese valves are expected to survive the spray experienced due to a crack in a moderate-energy line.

[ Wes!!!)ghouse 410.413-1

i m

i PSER Open item ,

i

Ihere are limited considerations for breaks and components in the turbine building. See subsections 3.6.L2.2 and 3.6.1.3.3 for additional discussion. Environmental qualification is not required for these considerations.

SSAR Revision:

In Table 3.11 1 note that the following valves and associated limit switches and solenoid valves are qualified for water spray fnom a moderate energy crack. (Add an "S" to the last column.)

CAS PL V014 CCS PL V200 CCS PL V208 CVS PL V047 CVS PL V090 PCS PL V001 A PCS PL V001H PCS PL V002A PCS PL V0028 PSS PL V0ll PSS PL V023 PSS PL V046 PXS PL'V042 RNS PL V011 RNS PL V022 SFS PL V035 SFS PL V038 VFS PL V003 VFS PL V010 VWS PL V058 VWS PL V086 WLS PL V057 WLS PL V068 Revise note 6 in Table 3.11 1 to add the follow.ng:

-I S = Qualified for operation with spray from a moderate-energy pipe crack or spray from a cold I high energy pipe crack.

E 410.413-2

4

-lNRC PSER OPEN ITEM {

m.n ;

n 4 16F (OITS.6391);

J Questior) 410.' . ,

- Re:

SSAR figures (A the VAS, VBS, VCS, VFS, VXS, and VZS are derived from AP600 piping and instrumentation diagrams (Pa!Ds), flowever, these figures do not accurately represent the details of the P&lDs, .

or the P&lDs and figures are not updated to reCect design changes. Additionally, Agures for other ilVAC; l

_ systems such as _VHS, VRS, and VTS are not provided. In order for the NRC staff to complete it: review of all.

1 the llVAC systems, Westinghouse must provide updated P&lDs and SSAR 6gures representing the current design. including major system instrumentation and system interactions, where applicable, such as supply of the -

' chilled and hot water being provided from the VWS and VYS.-

t

Response

Figures 9.4.1 1,9 4.21,9.4.31,9.4.61,9.431, and 9 4.101 are derived from the latest revisiors of the P&lDs for the VBS, VXS, VAS, VCS, VFS, and VZS systems respectively. Updated figures are not required.--- 4 The Health Physics and llot Machine Shop flVAC System (VHS) is described in SSAR 9.4.11. The Radwas:e '

' !!uilding !!VAC System (VRS) is described in SSAR 9.4.8.- The Turbine Building Ventilation System (VTS) is described in SSAR 9,4.9. Although these are a relatively simple systems, figures will be provided in the SSAR consistent with Westinghouse responses to Open items 410.362F,410.356F and 410.358F respectively.

Interfaces .with the Central Chilled Water System (VWS) and the Hot Water lleating System (VYS) is either presently shown in the SSAR or will be included with the addition of the new VHS, VRS and VTS figures as applicable.'

Instrumentation is show as appropriate in all SSAR figures.

- SSAR Revision:

.None-6 f

TBS" 410.416F 1 i

NRC F8ER OPEN ll:M P

m mass e 1

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- Question 410 419F (OITS 6394) _

f Re: !

1.- SSAR Section 9.4.1.1.1 states that those portions of the VBS which penetrate the MCR envelope are safety related and designed as seismic Category _1 to provide isolation of the MCR envelope from the '

surrounding areas and outside environment in the event of a DBA. Therefore, SSAR Figure 9.4.1 1 must be revised to_ show the piping separation with the required piping and component classifications including safety related pipe sizes.

2. SSAR Section 9.4.1.1.2 states that the VBS maintains MCR/TSC areas at a slight positive pressure with respect to the adjacent rooms and outside environment during normal operation to prevem infiltration of unmonitored air into these areas. liowever, Westinghouse needs to state exactly what negative pressure is required to be maintained in these r.reas with respect to the outside environment and adjacce. clean plant areas.-

. Response

1. SSAR Revision 18, Figure 9.4.1 1 (Sheet 5 of 7), shows the piping classifications and classificat ion changes. Component classifications will be clarified with changes to Figure 9.4.1 1 (Sheet 5 of 7) and Table 3.2 3 as shown below.

2. The MCR/fSC areas will be maintained at a slightly positive pressure as noted in the SSAR. A positive pressure exists because the supply air flow rate is maintained greater than the exhaust air flow rate. The value of the slight pressure is not a definitive parameter of the system design This is a nonsafety-related functmn For this reason, no change to the SSAR is necessary.

~SSAR Revision:

Figure 9 41 1 (Sheet 5 of 7) Note 2 will be deleted and replaced with the follswing:

2. The MCR pressure boundary HVAC isolation valves are constructed in accordance with ASME 111, Class J.

410.419F-1

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v NRC FSER OPEN ITEM . J!

a s -

_' - Table 3.2 3 (Sheet $4 of 67) Lines 8 through 13 will be revised as shown below; -

VBS.PleV1861 : MCR isolation Valve C1 ASME !!I.3 L ,

VBS PL VIR7 - - MCR ! solation Valve' C- 1 ? ASME !!! 3 >

VBS PL Vl88 ..MCR isolation Valve. C- I LASME lli 3 j VBS PL V189 - : MCR ! solation Valve 1 C. 'I ASME !!! 3 l

.VBS PL Vl90. MCR isolation Valve - C .- -1 ASME !!! 3 i

,. VBS PL Vl91- MCR isolation Valve - C 1- ASME !!! 3 .;

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NRC FSER OPEN ITEM

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Ouestoi '410.420F (OITS 6395)

Re:

RAI #: 410.420F Westinghouse needs to revise SSAR Section 9.4.1.2.2 to state that Se isolation dampers are tested in accordance with N5!0-1989, " Testing of Nuclear Air-Cleaning Systems, Standards."

Response

, SSAR subsection 9.4.1.2.2 will be revised to include leak testing in accordance with ASME N510.

SSAR Revision:

Subsection 9.4.1.2.2, revise to state:

' Isolation Dampers and Valves Nonsafety-related isolation dampers are bubble tight, single or parallel blade type, ne isolation dampers have spring return actuators which fail closed on loss of electrical power. De isolation dampers are constructed, qualified, and tested in accordance with' ANSI /AMCA 500 (Reference 14) or ASME N509 (Reference 2),

Section 5.9.

De main control room pressure boundary penetrations include isolation valves, interconnecting piping, and vent and test connection with manual test valves. The isolation valves are classified as Safety Class C and seismic l Category 1. Their boundary holation function will be tested in accordance with ASME N510 (Reference 3).

' The main control room pressure boundary isolation valves have electro-hydraulic operators. The valves are designed to fait closed in the event of loss of electrical power. The valves are qualified to shut tight against control room pressure."

410.420F-1 g

NRC FSER OPEN ITEM :

g;msmq

~ :

Queston 410.421F (OITS 6396)

Re:

RAI#: 410.421F.

Westinghouse needs to revise SSAR Section 9.4.1.2.2 to state that the MCR envelope isolation valves, isolation test ;

valves, and penetration test valves are ASMii Section Ill-Class 3.

Response

SSAR subsection 9.4.1.2.2 will be revised to indicate the ASME Section 111 class.

SSAR Revision:

Subsection 9.4.1.2.2, revise to state:

Isolation Dampers and Valves Nonsafety related isolation dampers are bubble tight, single or parallel. blade type, ne isolation dampers have spring return actuators which fait closed on loss of electrical power. The isolation dampers are constructed, qualified, and tested in accordance with ANSI /AMCA 500 (Reference 14) or ASME N500 (Reference 2),

Section 5.9.

The main control room pressure boundary penetrauons include isolation valves, interconnecting pping, and vent i and test connection with manual test valves, he isolation valves are classified as Safety Class C (sec

! subsection 3.2.2.5 and Table 3.2 3) and scismic Category 1.

De main control room pressure boundary isolation valves have electro-hydraulic operators. The valves are designed to fail closed in the event of loss of electrical power. He valves are qualified to shut tight against con.rol room pressure "

6 410,421 F-1

/

l; N: ; FSER OPEN ITEM 1.. .

Queshon 410 424F (OITS 6399)

Re:

The acceptable location of the single control room outside air intake serving the VBS conforms to the guidance of Section 6.4 of the SRP and RO 1.95 and was based on a previous SSAR figure (Figure 1.210) showing its location at elevation 153'-O' and previous SSAR Figure 1.217 which showed that it was located 35.1 m (115 ft) laterally and 26.2 m (86 ft) vertically below the plant vent discharge point. Ilowever, for the current design Westinghouse needs to verify the location and provide an appropriate figure for the air intake's with appropriate details as indicated here.

Response

Section 6 4 of the SRP provides guidance such that, * , the control room ventilation inlets should be separated from rnajor potential release points by at least 100 feet laterally and by 50 feet vertically." The SSAR will be revised as noted below. A figure is not required.

SSAR Revision:

Resise the first sentence in SSAR 9.4.1.2.1.1, second paragraph as follows:

"Outside supply air is provided to the plant areas sersed by the main control room / technical support center IIVAC subsystem through an outside air intake duct that is protected by an intake enclosure located on the roof of the auxiliary buildmg at elevation l53'-O' The outside air intake duct is located more than 30 feet below and more than 100 feet laterally awayfrom the plant vent discharge. The supply, return, and toilet exhaust i

u24m y wesuncous.

1 NRC FSER OPEN ITEM -

ummutmE

' Queston '410.431F (OITS 6406) i i

Re:

During abnormal operation, when high gaseous radiocctivity (111011) is detected in the MCR supply air duct and the MCR/TSC llVAC rWese is operable, both supplemental air filtration umts automatically start to pressurire the MCRffSC areas to at least 0.03 LPa (1/8 in water gauge) using filtered makeup. One of the supplemental filtration units is manually shutdown; the normal outside air makeup duct and the MCR and TSC

- toilet exhaust duct' isolation valves close; and the smoke / purge isolation dampers close,if open. The subsystem AllU' continues to provide cooling in the recirculation mode by maintaining the MCRE passive heat sink below Lits initial ambient air design temperature and maintaining the MCR/TSC areas within their design temperature.

The supplemental Gltratico pressurizes the combined volume of the MCR and TSC concurrently with filtered air.

1 A portion of the recirculated air from the MCR and TSC is also filtered for cleanup of airbome radioactivity.

llowever, Westinghouse needs to provide the following: ,

l. State that the positive pressurization of 0.03 kPa (1/8 in. water gauge) is "with respect to the adjacent rooms and outside environment" and revise the SSAR Section 9,4.1.2.3.1 accordingly.

2. The portion (percentage) of the recirculated air from the MCR and TSC that is filtered for cler.nup of

airborne radioactivity.

Response

1. SSAR 9.4.1.2.3.8 will be revised as requested, (See "SSAR Revision' below).

2.- 3SAR Table 9.4.1 1 (sheet I of 3) shows the nominal MCR/TSC subsystem Dow rate 22,000 cfm. S$AR Tabl: 9.4.1 1 (sheet 2 of 3) shows the nominal air filtration system Dow rate as 4,000 cfm. The portion of the recirculated air that is filtered is 4,000/22,000 or approximately 18 percent. No SSAR revision is required.

SSAR Revision:

1, Resise the first sentence in SSAR 9.4.1.2.3.1 second paragraph under Abnormal Plant Operation as follows:

If "high* gaseous radioactivity is detected in the main control room supply air duct and the main control room / technical support center liVAC subsystem is operable, both supplemental air filtration units automatically start t,o pressurize the main control room and technical support center a cas to at least 1/8 inch ug with respect ti che surrounding arens and the outside environment using filtered makeup air.

2. ' None 410,431F 1 g

NRC FSER OPEN ITEM Question 420.128F (OITS #6496) [

SSAR section 7.3.2.1 references WCAP 13594 ano WCAP 13022 (SSAR Section 7.2.4), FMEA of Advanced Passive Plant Protection System" as the Failure Mcde and Effects Analysis docketed information supporting the evaluation of the protection and safety monitosing system. The staff took credit for the FMEA analysis contained in these WCAPs in its safety evaluation. By letter NSD-NRC 5484, dated 12/11/97, Westinghouse responded to FSER open item 430.127F (OITS#6242) on the CMT level instrumentation in which a detailed FMEA is provided and used to Just!fy the CMT level instrumentation design arrangements. The staff concludes that the FMEA for the CMT tevel instrumerotion should be included in the FMEA WCAPs 13594 and 13662 as a condition of the statt's acceptance of the CMT levelInstrumentation design,

~ Response:

Attached is an Appendix for WCAP 13594 (Prop.) /13662 (Non Prop.) which includes the CMT level

. instrument FMEA. Note that this FMEA has a different purpose than the other FMEAs performod in this WCAP. The WCAP FMEAs were performed on processing hardware within the PMS, such as processing boards, l/O boards, etc. The CMT level instrument FMEA is performed on the piping I valving arrangement inorder tu evaluate issues associated with sharing level tap connections. As a result, the CMT level instrument has been placed in an appendix to this WCAP, A reference to these WCAPs is added to SSAR subsection 6.3.7.4,1.

SSAR Change:

Add referenco to these WCAPs in SSAR subsections 6.3.7.4.1 and 6.3.9.

ITAAC Change:

None gg 420.12aF-1

NRC FSER OPEN ITEM imimm a

Revision to SSAR subsection 6.3.7.4.1 6.3.7.4 Level Indication 6.3.7.4.1 Core Makeup Tank Level Ten 16 eel channels are installed on each core makeup tank. There are 2 wide range level channels which are used to confirm th&t the core makeup tanks are maintained at full water level during normal operation. There are four narrow range level channels which are used to control the actuation of the automatic depressurlZation system stage 1 valves. There are four narrow range level channels which are used to control the actuation of the automatic depressurization system stage 4 valves. Each wide range channet providos levet indication and alarms in the main control room. Each narrow range @.annelpavides discrete levet setpoints for indications and alarms in the main control room and for actuation of the automatic depressurization system. Each set of four narrow range channels share upper and lower level tap connections with the core makeup tanks; a failure modes and effects analysis confirms the ability of this arrangement to tolerate single failuies (Reference 2).

6.3.7.4.2 Accumulator Level Two level channels are installed on each accumulator. The level ind cations are used to confirm that accumulator levelis within bounds of the assumptions used in the safety analysis. Each channel provides level indication and both high &ad low level alarms in the main control room.

y w , ,,, no.i2sr-2

.l l

NRC FSER OPEN ITEM -

i

.i Revision to SSAR subsection 6.3.9 6.3.8 . Combined Liconse Intormation

~ 6.3.8.1 Containment Cleanliness Program The Combined License applicants referencing the AP600 will address preparation of a program to.

limit the amount of debris that might be left in the containment following refueling and maintenance -

outages.-

.6.3.9 References

1. WCAP 8966,

Evaluation of Mispositioned ECCS Valves," September 1977.

- 2. WCAP 13594 (Proprietary), WCAP 13662 (Nonproprietary), "FMEA of Advanced Passive Plant Protection System."

g 420.128F-3 l.

NRC FSER OPEN ITEM -

1 l

Appendix A to WCAP-13594 1.0 CMT LevelInstrument FMEA Evaluation Figure 1 shows the arrangement of differenpal pressure ;nstruments that are used to measure the core makeup tank (CMT) level. The two wide ringe DP transmitters provided on each CMT are nonsafety-related and are not further discussed in this evaluation. The eight narrow range DP instruments provided on each GMT are safety related and are the subject of this evaluation. The narrow range DP level instruments provide signals which Indicate the draining of the CMT during accidents and actuate automatic depressurization (ADS) valves. There are two ADS level setpoints in each CMT. The higher level setpoint actuales ADS stage 1 valves and the lower level setpoint actuates ADS stage 4 valvos.

Narrow range instruments are used for each ADS setpoint to provide acceptable measurement uncertainties associated with the harsh containment environment and the CMT water and reference leg temperature variations. As shown on figure 1, the four narrow range level instruments used for each ADS setpoint, share CMT level taps. Tne purpose of this Appendix is to demonstrate that credable single failures do not resutt in unacceptable multiple failures of the CMT level instruments.

1.1 FMEA Evaluation Each CMT has eight narrow range level switches. These switches do not have continuous readouts.

OP switches were selected because of their simplicity, reliability and use in similar operating conditions.

When the CMT water level drops below the ADS setpoint (as a result of CMT injection), a contact is closed in each level switch to provide indication and actuation.

Figure 2 shows additional details of one set of the narrow r' ue CMT DP level instruments including their cabbration valves. Credable failures include the misposn ; of valves and leaks in the sensing lines. Failures such as mispositioning of locked valves and pluggog of sensing lines are not considered credable single failures as discussed below.

Table 1 shows the failure modcs and effects analysis (FMEA) for this narrow range level switch arrangement. Discussion of this FMEA follows:

1. Inadvertent closure of one of the root isolation valves (V1 or V2) would block operation of all 4 DP instruments associated with that CMT connection. These valves are locked open to eliminate the need to consider these valves from being inadvartently closed as a credible single f ailure.

2. Inadverterit closure of one of the DP level sensor isolation valves (V3 or V4) would result in blocking one DP tovel sensor. This is considered a credible single failure which can be tolerated because there are 3 other DP sensors wnich can actuate ADS during an LOCA.

3. Inadvertent opening of one of the DP level sensor equalization valves (VS) would result in failure of the one CMT level sensor. The 3 other DP level sensors would be unaffected because the common g 420.128F-4 g

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upper line is required'to be located at the elevation of the upper level tap connection to the CMT.

As a result, as the CMT level drops below the upper level tap the 3 other DP level sensors would retain filled sensing lines up to the elevation of the upper iap. The 3 unaffected DP level sensors are sufficient to actuate ADS during an LOCA. ;

4. Inadvertent opening of one of the DP sensor vent valves (V6 or V7) would cause an RCS leak. It is not considered credible for this f ailure to occur during high pressure operation because the l instruments will not normally be maintained or calibrated in such conditions. Even if a operator. did start to open one of these valves he would hear / see the cod water leaking out as soon as he cracked the valve part way open and immediately close the valve. An error during shutoown maintenance could leave one of these valves open. However, as soon as the RCS is started to be pressurized the leakage from the RCS would be detect 9d whilo the plant is cold and at low pressures (when only one CMT is required).

5. Leakage through one of the DP senscr vent valves (V6 or V7) would cause an RCS leak. If this leakage is greater than 0.5 gpm it will be quickly identified and isolated. A leak of 0.5 gpm will produce a DP error of about 1 inch of water. Leakage through V6 will cause a lower level reading:

the error is smail enough that the ADS actuation setpoint would not be reached. Leakage through -

.V7 will cause a higher level reading; the error is small enough that the DP sensors will be able to actuate ADS as required should a LOCA occur.

6. Breaking of an upper level tap would cause the four DP sensors to read low and satisfy the 2 / 4 tow level logic portion of ADS actuation. However, actuation of ADS will not occur because the low level actuation signal is interlocked with the CMT actuation signal which will not be generated following such a small break because the CVS can makeup for the fluid loss through the 3/8" orifice in the CMT connection,

7. Plugging of an upper level tap would have similar effects to closing the upper tap root valve (V1).

This failure mode is not credible because the CMT has very good water quality and Decause the lines are verified to not be plugged during their calibration. Reactor coolant makeup is used to fill the CMT's and the inside surfaces of the CMT's and its connecting piping are stainless steel.

8, Breaking of a lower level tap would cause the four DP sensors to read high. This failure would prevent actuation of ADS from the affected CMT ADS is not required following such a small break because the CVS can maketip for the fluid loss through the 3/8" orifice in the CMT connection.

Even if ADS is requir9d the other CMT and its levelinstruments woukt provide the actuation signals even with consideration for a single failure.

9c Plugging of a lower level tap would have similar effects to closing the lower tap root valve (V2).

. This failure mode is not credible because the CMT has very good water quality. Reactor coolant makeup is used to fill the CMT's and the inside surfaces of the CMT's and its connecting piping are stainless steel, 420.128F-5

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= q n e..g i 1.2 FMEA Conclusions The FMEA evaluation of the CMT narrow range level switch arrangement shows that there are no credable single f ailures that can result in unacceptable multiple f ailures of these CMT level instruments.

1 (20.128F-6

NRC FSER OPEN ITEM Table 1 - FMEA et CMT Level Instruments Component Fr;ilure Effect on CMT Failure Detection Remarks Mode Valve V1 or closed Falls tour narrow range DP Administrative controls Locking these valves V2, normally level sensors. Other CMT on valve position locks. open climinates this open, locked will provide protection in failure mode open most LOCAs. Wide range DP level sensors will help operators to use manual ADS.

Valve V3 or closed Falls the one associated DP Calibration of V4 normally level sensor. The other 3 instrument every open DP level sensors are refueling outage will unaffected by this f ailure. detect valve mispositioning.

Valve VS, open No effect on DP level Calibration of normally closed sensors because common instrument overy upper sensor line is located refueling outage will at the elevation of the CMT detect valve i connection, mispositioning.

Valve V6. open Causes four narrow range The open valve results Because CMT's are normally closed DP sensors to read low, in increasing CMT teak located inside satisfying the ADS actuation as RCS pressure is containment, the most setpoint. ADS would be increased. RCS likely time this valve actuated if CMT's are leakage and CMT top would be actuated. temperature mispositioned is instruments would dunng shutdown detect this leakage, conditions. CVS has Leakage would be capability to makeup detected and isolated for leakage, before plant is put in conditions where a LOCA is possible, g 420.128F-7

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Component Failure Effect on CMT Failure Detection Remarks Mode leak Causes the affected DP RCS leakege and CMT CVS makeup is sensor to read low; below top temperature capable of maintaining the ADS actuation setpoint. instruments would RCS conditions such The other 3 DP sensors detect this leakage. that the CMT's do not would read only slightly low drain.

such that the ADS setpoint would not be not be satisfied as a result of the leak. ADS would be actuated it a LOCA occurred and the CMT's drained.

Valve V7, open Causes four narrow range Causes increasing CMT Because CMT's are-normally closed DP sensors to read high, leak as RCS pressure located inside which prevents them from is increased. RCS containment, the most actuating ADS. CMT level leakage and CMT top likely time this valve s9nsors in the other CMT temperature would be provides protection for non. instruments would mispositioned is LOCA events. detect this leakage. during shutdown Leakage wo.>ld be conditions. CVS has detected anu isolated capability to makeup before plant is put in for leakage.

conditions where a LOCA is possible.

leak Causes the affected DP RCS leakage and CMT CVS makeup is sensor to read high, such top temperature capable of maintaining that it would not reach its instruments would RCS conditions such ADS setpoint if the CMT detect this leakage. that the CMT's do not drained. drain.

The other 3 DP sensors would read only slightly high such that their ADS setpoint would be satisfied it a LOCA occurred and the CMT's drained.

420.128F-8 g ,

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Component Failure Effect on CMT Failure Detection Remarks Mode Upper Tap (L1) break Causes four narrow range The break results in a CVS makeup is DP sensors to read low, significant CMT teak. capable of maintainlng satisfying the ADS actuation RCS leakage and CMT RCS conditions such setpoint. ADS would be top temperature that the CMT's do not actuated if CMT's were instruments would drain.

actuated. detect this leakage.

The leak would be isolated before plant is put in conditions where a LOCA is possible.

plug Falls four narrow range DP During catioration the Calibration and water level sensors, Other CMT sensing lines will be quality eliminate this will provide protection in shown to be unplugged, f aibre mode, most LOCAs... Wide range Plugging is noi DP level sensors will help considered likely operators to use manual because the CMT water ADS, !s reactor grade water and the surf aces in contact with the water are stainless steel.

Lower Tap (L2) break Causes four narrow range The break results in a CVS makeup is DP sensors to read high, significant CMT teak, capable of maintaining which prevents them from RCS leakage and CMT RCS conditions such actuating ADS. CMT level top temperature that the CMT's do not sensors in the other CMT instruments would drain, provides protection for non- detect this leakage.

LOCA events. The leak would be isolated before plant is put in conditions where a LOCA is possible.

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NRC FSER OPEN ITEM Component Failure Effect on CMT Fa!!ure Detection Remarks Mode plug Falls four narrow range DP During calibration the Calibration and water level sensors. Other CMT sensing lines will be quality eliminate this will provide protection in shown to be unplugged, f ailure mode.

most LOCAs. Wide range Plugging is not DP level sensors will help considered likely operators to use manual because the CMT water ADS. is reactor grade water and the surf aces in contact with the water l are stainless steel.

420.128F10 W Westinghouse

NRC FSER OPEN ITEM i

Figure 1 CMT Level Instruments INLET e-- m T( ,,

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Figure 2 CMT Level Instrument Details er3 'o u

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f. CMT tewW se'tches have stam and ADS 6Ctuation setets 1 Locate p LevM tag tWw stout P belos CNT servwcm 420.128F-12 W Westingtmuse

NRC FSER OPEN ITEM er Question 420.129F (OITS #6497)

Westinghouse has included a note with the CMT level instrumentation drawing which states that the location of the upper and tower CMT level headers should be about 1 inch lower than their connecticn to the CMT, This is an important design feature which should be included for verificatior. in .lTAAC.

Response

An ITAAC will be added to verify the elevation of the CMT narrow range levelin;trument upper level tap line. The elevation of the lower header is not safety importt.nt and does not need to be verified in ITAACs.

1AR Change:

Nonc ITAAC Change:

Attached is a revision to ITAAC Table 2.2.3 4 g 420.129F-1 r: _

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Table 2.2.3-4

- Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria -

8 c) The PXS proviaes RCS v) Inspections of each of the v) The calculated volume of makeup, boration, and safety following tanks will be each of the following tanks is as injection during design basis conducted: follows:

emits. -

CMTs -

CMTs 2 2000 ft' Accumulators - Accumulators 2 2000 ft' IRWST -

IRWST a 557,000 gal between the tank outlet connection and the tank overflow vi) inspection of the as-built vi) Plates located above each components will be conducted containment recirculation screen for plates located at,ove the are no more than 10 ft above the containment recirculation top of the screen and extend out screens. at least 10 ft from the trash rack portion of the screen.

vii) Inspections of the IRWST vii) The screen surface area and containment recirculation (width x height) of each screen is screens will be conducted. 2 70 ft'. The bottom of the containment recirculation screens is 2 2 ft above the loop compartment floor. _

viii) Inspections will be viii) The type of insulation used conducted of the insulation on these lines and equipment is used inside the containment not a liberous type.

on ASME class 1 lines and on the reactor vessel, reactor coolant pumps, pressurizer and steam generators.

lx) Inspections will be Ix) The centerline of each conducted of the CMTlevel upper level tap line at the tee sensors (PXS-11NB/D/C, - for each level sensoris 12NB/C/D, 13NB/C/D, - located 1"21" below the 14NBIC/D) Upperlevel tap centerline of the upperlevel lines. tap connection to the CMT.

g , 420.129F-2

NRC FSER OPEN ITEM Question 440.753F (OITS 641?)

Increase in the RCS Inventory Esents (SSAR 15.5)

In on NRC inspection of Westinghouse AP600 design control activities from Nosember 17 through 2h 1997, the staff found that operator actions were necessary (opening of the reactor vessel head sents) to present over611 of the pressurizer under some of the conditions in the analyses of increased RCS insentory events. The limiting case presented in the SSAR states that it " bounds cases that model explicit operator action 30 minutes after reactor trip." It is the stafi's understand ing that the cases that model explicit operator action actually require operator action or the pressurizer oser611 would be worse than reported in the SSAR. It appears that Westinghouse has reported the worst cverGil transient tha' ioes not require operator action but not the worst transient if no operator action were assumed.

Consequently, the staff concludes that operator action is necessary to mitigate the worst increase in RCS inventory events. Westinghouse should proside the following additior.sl information relative to the cases that model evflicit operator action within 30 minutes:

(a) Discuss assumptions important to the calculations related to pressurizer over611 resulting from increased RCS inventory events where operator action is necessary. What is the maximum time delay that can occur without taking operator action before the transient would exceed the limits currently reported in the SSAR.

(b) Proside information to demonstrate that unambiguous alarms or indications for the events are available, and the pocedural instructions are clear to operators to take appropriate actiora within the time frame assumed in the analyses (c) Westinghouse stated that the cases that model explicit operator action take credit for the use of the reactor sessel head sents to reduce the RCS inventory. Address compliance of this case with the technical specification requirements of 10 CFR 50 ~,6. Specifically, item (c)(2)(ii)(C), criterion 3 for the TS requirements, states that "A structure, system, or comoonent that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier."

(d) Update the SSAR to include the limiting analyses that credited the operator actions to present pressurizer overfill from occurring.

(c) Discuss the need to add an ITAAC to verify the capacity of the reactor vessel head vent system used by operators to prevent pressurizer over611 from occurring.

Response

(a),(b) SSAR Section 15.5 has been revised to include a more complete discussion of the use of operator actions to mitigate the consequences of the increase in reactor coolant inventory W53N T Westinghouse

c

[

i NRC FSER OPEN ITEM esents discussed in this RAI. The resised discussion addresses the assumptions important to i the calculations related to pressuriier oserfill resulting from increased RCS inventory esents t where operator action is necessary and addresses thetasis for the timing of the operator i actions credited it also includes a discussion of ths instrumentation used by the olerator, i t

n (c) Per the response to FSER Open item 440.78$. a technical specincation for the reactor sessel ;

' head vent valscs has been incorporated. l (d) The resised SSAR 3ection 13.$ discusses the analyses of events where operator actions are }

assumed flowever, the results of these esents are less limiting (the cargin to pressunter ;

osern!!is greater) than the analyses presented in thu SSAR, and the annipes presented in the

)

SSAR are bounding.

(c)- The RCS ITAAC (DM 2.1.2) currently contains design committments for the reactor sessel head vent vahes reg vding their safety class and active safety funw.m. It is modified as l shown to address the required capacity to accomodate overfill scenarios. !

6 SS/.R Reglelent i

3SAR Sution l$.5 !

t ITAAC Resisloret l The followh.g design cornmittement will be added to the RCS ITAAC and CDM Table 2.1.2 4:

i lhe I:CS provides emergemy letdown during design basis events fhe follo ~ Inspection, Tests and Analpes will be added to the CDM Table 2.1.2 4: l Inspectrons of the reactor vemt head vent vahrs and Inlet and outlet piping will be conducted. ?

The following Acceptance Criteria will be added to CDM Tame 21.2 4:

A report extsts and concludes that the rapactry of the reactor vess0 head vent Is sufflctent to pass not ;

' n than K2 lbu%ee at 1250 psia in the RCS i

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15. Accident Analpte 15 S.1 3 Results Figures 15.5 l 1 through 15 5.l 11 show the transient response to the inadsettent operation of the core makeup tanks dunng power operation. The inadsenent opening of the core makeup tank discharge salses occurs at 10 seconds Inhibited core makeup tank injection begins as the reactor coolant pumps continue te opante. The pressuriier lesci initiall>

shrinks due to the addition of cold borated water. As the core makeup tanks continue to add insentory to the primary s> stem, the pressuriier lesel beings to increase until the high 3 pressuriier Itvel reactor trip setpoint is reached at about 1599 seconds. After a 7 second delay, the neutron aus starts decreasing due to the reactor trip, which is immediately followed by the turbine trip. Following reactor tnp, the reactor power drops and the ascrage reactor coolant system tempeinture decreases with subsequent coolciut shrinkage. However, due to the assumed loss of offsite power, the reactor coolant pumps trip at about 1604 seconds and the core makeup tanks start injecting cold water into the reactor coulant system at a much higher rate. The pnmary coolant system shrinkage is counteracte:i by the core makeup tank injection, and the pressuriier uater solume starts to increase because of the heatup of the cold injected Guid by the decay heat. The high 3 pressuriier leve! setpoint it once again reached at about 1653 seconds, and after a 15 second delay, the signal is tent to actuate the PRilR heat exchanger and block the pressuriier heaters, Following a consenatise 22 second delay, the utses ve asumed to open to actuate the PRiiR heat exchanger at about 1690 secot.ds.

After reactor trip, the pressure in the pnmary and secondary systems increases initially due to the assumed unavailability of the nonsafety related control systems = The primary and secondary system pressures esentually decrease as the PRilR system remoses decay heat.

The core makeup tanks uork in recirculation mode, meaning they are alway s alled *cith w ater because cold borated water injected through the injection lines is replaced by hot water coming from the cold leg (balance lines). The pressurizer lesel increases until the core makeup lank reci culation is decreased sufficiently, and the PRHR system heat removal rate approaches that of the core deca) heat ymeration.

At about 6,827 seconds, the PRiiR heat Hus approaches core decay heat and the pressuriier water solume stops increasing. At 6,966 seconds, the pressunier safe s alses close. At approximately 10,000 seconds, the PRiiR heat aus ir.atches the core decay heat. At approsimately 18,000 seconds, the core makeup tanks essentially stop recirculating.

Figure 15 5.16 shows the departure from nucleate boiling ratio (DNBR) until the time of reactor coolant trip and subsequent now coastdown due to the loss of oITsite power. At this time, core power and heat Ous base diminished sufficiently, due to the reactor trip, that DNBR is well above the design limit salue denned in Tection 4.4.

The calculated sequence of esents is shown in Table 15.51.

1 The limiting case presented here bounds all cases that model explicit operator action i 30 minutes aner reactor inp. For such cents, the operator would take action to redue: the i increase in coolant inventory. As the pressunier water level would increase above the high I pressuriier water level that nonnally isolates chemical and* volume control system makeup, Hesisinn: 13 May 30,1997 l3,3 4 Westingn0USS Wo.753P3

l$. AtridsCt ACalpes I the nonn.sl letdown line could be placed into senice to reduce the increase in coolant I ins entory. If letdown could not be placed into senice, the operator could use the safety I related reactor sessel head unt vahes to rc '9ce the increase in coolant insentory. For these i esents,following the procedures outlined in tne AP600 Emergency Response Gudelines AFR-l 1.1, there is sufficient time for the operator to mitigate the consequences of this esent, and the I results of such an esent have a greater margin to pressuriier oserfill than that presented in I this analysis.

15.5.1.4 Conclusions The-lim it in g -+ase-presated-her e4ound e-ell-eaw +-t hat-modem plicit -oper at or-aotion

.h in u te,-e ft waeactor-t r i p-Con sid ering-4 hiti The r suits of this analysis st ow that inadscrtent operation of the core makeup tanks during power operation does not adscrsely affect the core, the reactor coolant system, or the steam system. The PRilR heat removal capacity is such that reactor coolant water is not reliesed from the pressuriict safety vahes.

DNDR always remains abose the design lirnit salues, and reactor coolant system and steam geacrator pressures remain below 110 percent of their design salues 15.5.2 Chemical and Ynlume Control S) stem Malfunction That increases Reactor Coolant Ingentory 15.5.2.1 Identification of Causes and Accident Description An increase of reactor coolant insentory, which results from addition of cold unborated water to the reacter coolant system,is analyred m subsection 15.4.6.

In this subsection 15.5 2, the increase of reactor coolarit sy stem insentory due to the addition of borated water is analyzed The increase of reactor coolant system coolant intentory may be due to the spunous operation of one or both of the chemical and solume control system pumps or by the closure of the letdown path. If the chemical and solume control system is injecting highly borated water into the reactor coolant system, the reactor esperiences a negatisc reactisity excursion due to the injected boron, causing a decrease in reacter power and subsequent coolant shrinkage.

The load decreases due to the effect of reduced steam pressure after the turbine throttle valve fully opens.

At high chemical and solume control system boron concentration, low reactisity feedback conditions, and reactor in manual rod control, an "S" signal will be generated by either the low T m or low steamline pressure setpoints before the chemical and solume control system can inject a signi6 cant annount of water into the reactor coolant system. In this case, the chemical and volume control system malfunction esent proceeds similarly to, and is only slightly more limiting than, a spurious "S" signal esent. If the automatic rod control is modeled and the pressunier spray functions properly to prevent a high pressure reactor trip signal. no "S" signals are generated and this specific esent is terminateu by automatic Resision: 13 W~

Westinghouse 15.5 5 May 30,1997

$ /o. 7 G 3 F 'l

l$. Atrident Anatpes The pressuriier heaters are automatically blocked on an *S" signal, and do not at heat to the system during the penod of fluid thennal expansion that produces the peak pressuriier water solume. Thus, the pressuriier heaters are assumed to be inoperable during this egent.

Pressuriier spray The spray s> stem controls the pressuriier pressure so that a high pressuriier pressure reactor trip is prevented.

130ron injection After 10 seconds at steady state, the chemical and volume control system pumps start injecting borated unter, which is slightly abose the reactor coolant system boron concentratiori. Upon receipt of an "S" signal, the chemical and volume control system pumps are isolated and the core makeup tanks begin injecting 3400 ppm borated water

Turbine load The turbine load is assumed constant until the governor dnves the throttle valse wide open Then the turbine load drops as steam pressure drops

Protection and safety monitoring s> stem actuations if the automatic rod control s> siem is modeled and the pressurarer spray system functions properly, no reactor inp signal is expected to occur. Instead, the esent is terminated by automatic isolation of the chemical and volume control r3 stem on the safety grade high 2 pressuriier les el setpoint. If the automatic rod control systern is not actise and the pressuriier spray sy stem is assumed to be available, reactor trip may be mitiated on either low T,a "S" or a low steamline pressure "S" signal.

The core decay heat is remosed by the PRilR heat exchanger. The worst single failure is assumed to occur in he outlet line of the PRiiR heat exchanger One of the two

,arallel isolation valses is assumed to fail open.

Plant systems and equipment as ailable to mitigate the effect of the accident are discussed in subsection 15.0.8 and listed in Table 15 0 6. No single actise failure in any of these systems or equipment adscrsely alTects the consequences of the accident.

15.5.2.3 Results Figures 15 5 21 through l$.S.212 show the transient response to a chemical and solume control s> stem malfunction that results in an increase of reactor coolant system inventory.

Neutron flus slowly decreases due to boron injection, but steam flow does not decrease until later in the transient when the turbine throttle valses are wide open Resisiont 13 May 30, IM 15 $ 8 3 Westingh0056 Wo.753F-f

ts, Accident Analpes As the chemical and solume control sy stem injection now increases reactor coolant system intentory, pressuriier water solume begins increasing while the pnmary sy stem is cooling down At about 1.236 seconds, the low Tm setpoint is reached, the reactor trips, ar d the control rods start musing into the core.

Immediatcly following reactor trip, the turbine is tripped and after a 3 second delay, a consequentialloss of offsite power is assumed and the reactor coolant pumps inp. The basis for the 3 second delay is desenbed in subsection l$ 0.14. Soon after reactor trip, the pressuriier heaters are blocked and the niain feedwat lines, steam lines, and chemical and solume control system are isolated After a consenntne 22 second delay, the PRilR heat eschangei !s actuated and the core makeup tank discharge satses are opened. The core makeup lanks work in recirculation mode, meaning they are alwsys Olled with water because cold borated water integrated through the injection lines is replaced by hot water coming from the cold leg balance lines The operation of the PRilR heat eschanger and the core makeup tanks cools down the plant.

Due to the swelling of the core makeup tank water, the pressunier lesel is still increasing.

At about 3,000 seconds, reactor coolant system ternperature is 455'F. The cooling effect due to the core makeup tanks is decreasing in this condition, the PR11R heat eschanger cannot remose the entire decay heat. Reactor coolant system tempe*sture tends to increase until an equilibrium between deca) heat power and heat absorbed by the PR11R heat eschanger is reached At approsimately 21,700 seconds, the PRilR heat flux approaches core decay heat, the pressuriier water solume stops increasing, and the pressunier safety salses close. At approsimately 22,000 seconds, the PRilR heat Ous matches the core decay heat and the core malcup tanks essentially stop injecting Figure 15 5 2 6 shows the DNBR until the time of reactor coolant pump trip and subsequent Oow coastdown due to the loss of olhite power. At this time, core power and heat nux hase diminished suf0ciently, due to the reactor trip, that DNBR is well above the design limit 5alue denned in Section 4.4.

The calculated sequence of esents is shown in Table 15.51.

I The limiting case presen'cd here bounds all cases that model esplicit operator action i hour i after reactor inp. Fr neh esents, the operator could take action to reduce the increase in I coolant insentory. As tb prest.utiier water lesel would increase abose the high pressuriier I water lesel that normally isolates chemical and solume control system makeup, the nonnal I letdown line could be placed into senice to reduce the increase in coolant inventory. If I letdown could not be placed into senice, the operator would use the safety related reactor I sessel head sent valses to reduce the increase in coolant insentory. For these esents, I following the procedures outlined in the AP600 Emergency Response Guidelines AFR l.1, I there is sufficient time for the operator to mitigate the consequences of this esent, and the I results of such an esent hase a greater margin to pressuriier oser0ll than that presented in I this analy sis.

Resision: 13 W Westingh00St 1339 May 30,1997

~

40 0. 7 5 3F %

l$. Arrident Ansipes 15.5.2,4 Conclusions l he limising +ase presented 4iere bound +ase*4 hat modeleplM6+perator waion-i-hour-efter reactor-inp-C+nsidering 4 hie,-The results of this analysis show that a chemical and solume control sptem rnalfunction does not adscrsely affect the core, the reactor coolant 53 stem, or the steam system. The PRHR heat remosal capacity is such that reactor coolant water is not reliesed from the pressuriier safety s als es. DNBR remains abos e the design limit s alues, s.nd reactos coolant system and steam generator pressures remain below 110 percent of their de.ign 5 alues.

If the automatic rod control system and the pressurir.cr spray systems are assumed to function, no reactor inp signal is espected to occur. Instead, the esent is terminated by automatic isolation of the chemical and solume control s3 stem on the safety grade high pressuriier les el setpoint. If manual rod controlis assumed and the pressuriier spray system is assumed to be unavailable, reactor trip may be initiated on either a high pressuriier pressure, low T , *S",

or a low steamline pressure "S" signal.

15.5.3 Iloiting Water Remetor Trm 44 1his subsection is not applicavic ,& & AP600.

15.5,4 Combined 1.icense Information This subsection has no requirement for additional information to be prosided in support of the Combined License application.

15.5.5 References I ilurrett, T W. T., et al.,"LOFTRAN Code Description," WCAP.7907.P.A (Propnetary) and WCAP.7907.A (Nonproprietary), Apnl 1984.

b Resision: 13 Ma) 30,1997 l3 31o 3 Westinghouse Wo. 75 3F4

i pMIRHij NRC FSER OPEN ITEM ! 3

^

Question 440.770F (OITS 6344)

Similar to the concern identined in 440 769F. the analysis of CVS malfunction is performed with the _

plant initially in hiode 1, full power conditions and is discussed in SSAR 15.5.2. The sesenty of the l csent depends on the Guid expansion, which,in turn,is controlled by the integral now from the CVS, j decay heat lesels, and heat rearm al from the PRilR 111 The stafT notes that in the absence of anal)ses to quantify the total effect of these parameters on Guid expansion,it :s not clear that the full.

power case discussed in SSAP.15.5.2 bounds esents initiated from operations below Mode 1.

Westinghouse is requested to malyic the CVS malfunction events at shutdown modes and show that the results are acceptable.

Responset An analysis of an increase in RCS inventory due to a CVS malfunction is provided in the SSAR subsection 15.5 2. This analysis is performed for the plant at full power conditions, and the results of this analysis conservatively bound these events that can occur during lower modes as discussed below.

A CVS malfunction is postulated as the inadscrtent operation of I or 2 makeup pumps to Gil the RCS.

Operation of the CVS makeup pumps is controlled by the plant control system. A single pump automatically starts on low (normal) pressuriier wster level. The control system is programmed to operate the pump to add makeup to the RCS at the prevailing kCS boron concentrahon. A single !

makeup pump operates and Gils the RCS to the high (normal) pressurizer water level. If the control system fails, and the makeup pump coniiracs to operate, the pressurizer water levi;l will increase until the high 2 pressuriter water level is reached. Upon receipt of a high 2 pressurizr.t wster level signal, :

the Protection and Safety Monitoring System (PMS) provides a safety related signal to close the CVS makeup isolation valses. This terminates the inadvertent makeup, and thereby terminates the event without a reactor trip or safeguards actuation ('S') signal.

While the abose scenario is the most probable outcome of a CVS malfunction, sescral combinations of makeup boron concentration, feedback conditions, and plant system in'.cractions hase been identined which can result in more limiting scenarios. For these scenarios, the CVS malfunction can cause a slight boration of the RCS. Subsequently, the core power decreases which then results in the generation of an 'S' signal on low cold leg temperature. The 'S' signal is generated prior to the pressuriict water lesel increasing to the high.2 pressuriter level signal used to isolate CVS makeup.

This results in the actuation of the CMTs and PRilR, and is discussrd fully in SSAR subsection 15.5.2.

These low probability scenarios are not a concern in shutdown modes. In shutdown modes, the CVS malfunction esent proceeds similarly to the full power cases discussed in the SSAR that do not result in an 'S' signal being generated on low cold leg temperature. The CVS malfunction results in the pressuriier water lesel increasing to the high 2 lesel setpoint, and the CVS makeup valves are then closed by the PMS. For these cases, an 'S' signal is not actuated. The reactivity aspects of the CVS malfunction that cause an 'S' signal for the at power cases does not occur at shutdown because the core is shut down. Boron concentration mismatch due to the CVS malfune: ion does not cause a s

440,770F.1 g

f

,+ .~. -. ,ww , - - r - - , , ,- , , . v a , ,m,- ,w- --

NRC FSER OPEN ITEM decrease in core power or cold leg temperature. Therefore, this esent is tennineted in all shutdown mode cases by the closure of the CVS makeup line isolation sahes by the PMS.

SSAR Resition:

None 0,77062 T Westinghouse

NRC FSER OPEN ITEM Question 440.783F (OITS #636?)

LCO 3.4.12 ADS . Operating LCO 3.4.12 specifies that the ADS, including 10 flow paths, shall be operable during MODES 1 through 4 operation, if one flow path, or one stage 1 flow path and one stage 2 or stage 3 flow path is inoperable, Action A.1 requires restolation of these flow path within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . BASES 3.4.12 does not provide sufficient justification for the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action completion time. In response to RAI 440.671 (Westinghouse letter, B. McIntyre to T. Ousy, NSD-NRC 97 5278, August 27,1997), Westinghovce provided justifications for the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time. Westinghouse contended that the basis for 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time is consistent with the STS PORV,3.4.11, Action B, and consistent with two train ECCS systems that can perform their safety function without a single failure. The applicant also contends ihat the inoperable ADS flow path conditions (Condition A) have been assumed as single failures in the Chapter 15 LOCA analyses. The staff finds that the design bases and functional requirements are so different between the PORVs and the AP600 ADS valves that it is not prudent to draw an equivalence between thern. In addition, a single failure assumption in the safety analyses is not to be used for a failure that has been found to exist. The applicant should provide additional justifications, which should be docume'1ted in the TS BASES, to justify the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time.

This is open item.

Response

The justification for allowing this condition to exist for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is that this time is long enough to fix many problems, the time is short enough that tho probabilt . I having a DBA is low and if a DBA did occur st can be rnitigated as long as a single failure does are also occur. Note that with ADS valves INOPERABLE in accordance with LCO 3.4.12 A (cne e % stage 1 and one ADS stage 2/3 paths) or LCO 3.4.12 B (one ADS stage 4 path), the AP600 cs mitigate any DBA as long as it does not suffer a single failure dunng the DBA. h additic.n, as mentie ed in the background for this LCO, the PRA shows that adequato core cooling can be provided with the failure of two (or more) flow paths.

The NRC guidance on technical specifications is that, "in consideration of the current AP600 review schedule, the completion times and surveillance frequencies for the AP600 TSs should be based on STS values. Where the AP600 design lias no equivalent STS system, we suggest that the times be based on the STS treatment of the comparable safety function.' For the ADS there is no equivalent STS system, however since the ADS supports safety injection, the ECCS STS may bo used to justify this Completion Time, in tne STS, an example of the safety injection safety function is provided by LCO 3.5.2, ECCS -

Operating. Condition A provides the Required Action for a level of degradation in which the minimum safety injection function is retained, such that any DBA can be mitigated, provided a single f ailure does not occur dunng the accident. The Completion Time for STS LCO 3.5.2 is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

g 440.783F-1

NRC FSER OPEN ITEM The following wording is proposed to clarify the justificatioq of the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time:

"A Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable based on the capability of the remaining ADS valves to perform the required safety functions assumed in the safety analysis and the low probability of a DBA during this time period. This Completion Time is the same as is used for oegraded ECCS systems which are capable of performing their safety function without a single failure."

A revision to this technical specification BASES is attached.

SSAR Change:

A revision to LCO BASES 3.4.12 is attached.

ITAAC Change:

None

[ We51}Dgh00$6

l NRC FSER OPEN ITEM BASES [LCO 3.4.12)

APPLICABILITY in MODES 1,2,3 and 4 the ADS must be OPERABLE to mitigate the potential consequences of any event which causes a reduction in the RCS inventory, such as a LOCA.

The requirements for the ADS in MODES 5 and 6 are specified in LCO 3.4.13, " Automatic Depressurization System (ADS) Shutdown, RCS Intact,"

and LCO 3.4.14, " Automatic Depressurization System Shutdown, RCS Open.'

l ACTIONS Al if any one, or if two flow paths, consisting of one stage 1 and one stage 2 or 3, are determined to be inoperable, the remaining OPERABLE ADS flow paths are adequate to perform the required safety function as long as a single failure does not also occur. A flow path is inoperable if one or two of the ADS valves in the flow path are determined to be inoperable. A Comp?st;ca Time cf 72 hcurs is accept &ble 3;nce the OPERABLE ADO psths con m;t,ge:e DCAs m hout & sing's f a;lure. A Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable based on the capability of the remaining AD3 valves to perform the required safety functions assumed in the safety analysis and the low probability of a DBA during this time period.

This completion time is the same as is used for two train ECCS systems which are capable of performing their safety function without a single failure.

.B.1 and B.2 If the Required Actions and associated Completion Times are not met or the requirements of LCO 3.4.12 are not met for reasons other than Condition A, the plant must be brought to MODE 5 where the probability and consequences on an event are minimized. To achieve this status, the pic.nt must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Tne allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner, without challenging plant systems.

d'o 783'-3 W w.mneus.

NRC FSER OPEN ITEM Question 440.788F (OITS #6445)

The staff recognizes the importance of estabbshing a process for ensuring that the performance of the actual ADS valves in an AP600 plant meets functional requirements consistent with those determined from the design certification test program and reflected in design basis analyses performec for the plant.

Accordingly, the staff has determined that the ADS " road map" documented in Westinghouse letter NSD NRC 976 5100, dated April 30, ',997, should be incorporated into the AP600 SSAR. In addition, the steps in the " road map" leading from the design certification test program to the qualification of the actual AP600 valves should be incorporated into the inspections, tests, analyses, and acceptance criteria (ITAAC) for the AP600, and cross referencing between the ITAAC, SSAR, and other appropriate documentation should be included, to ensure that the process is properly and consistently implemented.

This is an Open item.

Response

The AP600 SSAR identifies the type of valv9s that will be used in the ADS. The portions of the ADS roadmap that are important to ensure that the ADS valves will be reliable int,lude:

- ADS valve quakfication

- ADS valve production operational verification

- Pro. operational valve operational venfication

- In service valve operational venfic.ation ADS valve qualification is addressed in SSAR subsection 5.4.8.1.2 (MOVs) and 5.4.8.1.3 (Other Power.

Operated Valves including Squibs).

ADS valve production operational verification will be performed. These tests are vendor specific and f

are not included in the SSAR.

ADS valve pre operational valve operational venfication is addressed in SSAR subsection 14.2.9.1.

ADS valve in. service valve operational venfication is addressed in SSAR subsection 3.9.6.2.2 and SSAR Table 3.916.

ITAACs are provided for the ADS valves in section 2.1.2. SeveralITAACs are provided. The specific ITAAC that relates to this RAIis item 12.a in Table 2.1.9 4. This ITAAC requires type tests for MOVs and a venfication that the installed valve is bounded by the type tests. There are no ITAAC requirements for type tests for the ADS squib valves. An ITAAC is proposed to require a similar type test for the ADS stage 4 squib valves.

g 440.788F-1

NHC FSER OPEN ITEM pgl SSAR Change:

None ITAAC Change:

Attached is a revision to the ITAAC Design Description 2.1.2 and ITAAC Table 2.1.2 4, item 12.a.

gg 440.788F 2 3,

l NRC FSER OPEN ITEM l

Revision to ITAAC 2.1.2 Design Description

8. The RCS provides the following safety related functions:

al The pressurizer safety valves provide overpressure protection in accordance with Section 111 of the ASME Boiler and Pressure Vessel Code.

b The reactor coolant pumps (RCPs) have a rotating inertia to provide RCS flow coastdowq on loss of power to the pumps.

4 c The RCS prov! des automatic depressurization during design basis events.

9. Ti:e RCS provides the following nonsafety related functions:

a) The RCS provides circulation of coolant to remove heat from the core, b) The ACS provides the means to control system pressure.

10. Safety related displays identified in Table 2.1.21 can be retrieved in the main control room (MCR).

11. a) Controts exist in the MCR to cause the remotely operated valves identified in Table 2.1.21 to perform active functions.

b) The valves identified in Table 2.1.21 as having protection and safety monitoring system (PMS) controt pertoim an active safety function after receiving a signal from the PMS.

c) The valves identified in Table 2.1.21 as having diverse actuation system (DAS) control perform an active safety function after receiving a signal from DAS.

l 12. a) The motor operated valves and the squib valves identified in Table 2.1.21 perform an active safety related function to change position as indicated in the table.

b) After loss of motive power, the remotely operated valves identified in Table 2.1.21 assume the indicated loss of motive power position.

440.788F 3 g

NRC FSER OPEN ITEM Table 2.1.2-4 (Cont.)

Inspections. Tests, Analyses, and Acceptance Criteria Design CommHment inspections, Tests, Analyses Acceptance Criteria 11.c) The valves I) Testing will be performed on the 1) The squib valves receive a identified in Table 2.1.21 squib valves identified in Table 2.1.2 signal at the valve electrical as having DAS control 1 using real or simulated signals into leads that is capable of actuating perform an active safety the DAS without stroking the valve. the squib valve.

function after receiving a signal from DAS. 11) Testing will be performed on the <

other remotely operated valves li) The other remotely operated identified in Table 2.1.21 using real valves identified in Table 2.1.21 or simulated signals ;nto the DAS. as having DAS control perform the active function identified in the table after receiving a signal from DAS.

12.a) The motor operated I) Tests or type tests of motor. 1) A test report exists and l valves and squ!o valves operated valves will be performed concludes that each motor-identified in Table 2.1.21 that demonstrate the capability of the operated valvo changes position perform an active safety. valve to operate under its design as indicated in Table 2.1.21 related function to change conditions, under design conditions.

position as indicated in the table, 11) Inspection will be performed for ii) A report exists and concludes the existence of a report verifying that the as installed motor.

that the as installed motor-operated operated valves are bounded by valves are bounded by the tests or the tests or type tests.

type tests.

lil) Tests or type tests of squib lil) A test report exists and valves will be performed that concludes that each squib demonstrate the capability of the valve changes position as valve to operate under its design Indicated in Table 2.1.21 conditions, under design conditions.

IV) Inspection will be performed lv) A report exists and for the existence of a report concludes that the as-Installed veritying that the as installed squib valves are bounded by squib valves are bounded by the the tests or type tests, tests or type tests, 440.788F 4 g

NRC FSER OPEN ITEM mummr Question 440,793F (OITS #6443)

Several important parameters of the PXS systems, which affect passive system performance and are j input parameters to the safety analyses of the design basis events, are not specified in the SSAR.

These include the following:

(a) The elevation of passive core cooling components relative to the RCS loops, including CMTs PRHR Heat Exchanger IRWST Injection Sumps Containment Recirculation Sumps (b) The piping flow resistances (including valves and spargers where applicable) associated with the CMis PRHR Heat Exchanger l IRWST injection Containment Rectreulation ADS stages 1,2,3, and 4 Accumulators (c) Flow area of she ADS Spargers l These parameters are inputs to the safety analyses of the design basis transients and accidents to demonstrate the PXS mitigation capability. The values of these parameters with allowance for uncertainties should be also be specified in the SSAR. This is open item.

Response

Detailed component, system, and structure parameter inputs to the safety analysis are not provided in the SSAR. They are provided through OA documents outside of the SSAR. Note that some plant parameters that are important to understanding the operation of the systems are provided in the SSAR, alttiough these numbers are not necessarily bounding values. The following identifv the parameters that are contained in the SSAR.

(a) SSAR subsection 6.3.6 contains the minimum CMT and IRWST elevations. Tlie elevation of the PRHR HX is not provided in the SSAR, however its natural circulation heat transfer rate is provided in SSAR Table 6.3 4, The location of the IRWST and containment recirculation screens are discussed in SSAR subsection 6.3.2 2.7, t.ithough specific locations are not provided.

(b) The SSAR does not provide these line resistances. Note that the PRHR HX natural circulation heat transfer rate is provided in SSAR Table 6.3 4.

"U #

T westinghouse

NRC FSER OPEN ITEM ij!!MiFp 1-(c) Tt,o response to RAI 640.113F, revision 1, adds the sparger flow area to SSAR Table 6.3 4.

SSAR Change:

None ITAAC Change:

None 440.793F 2

i NRC FSER OPEN ITEM 1

Question 440.804F (OITS #6478)

Soction A of SECY 94 084,

Policy and Technicalissues Associated w!th the Regulatory Treatment of Non Safoty Systems (RTNSS) in Passive Plant Designs," March 28,1994, discusses the process used (a) to develop hsights regarding the importance of non safety related systems to the overall safety of the AP600 design, and (b) to determine what, if any, additional regulatory controls should be implemented for those non safety related systems determined to be important to safety. Chapter 22 of the FSER discusses the RTNSS process in detall.

Westinghouse's original evaluation of RTNSS implementation is discussed in WCAP 13856, *AP600

!mplementation of the Regulatory Treatment of Nonsafety Related Systems Process? In addition, the focused PRA sensitivity study that forms a major part of the RTNSS process is contained in Chapter 52 of the AP600 PRA. The original evaluation in WCAP 13856 identified only two conditions requiring regulatory controls on non safety related systems: the reactor trip functior, of the Diverse Actuation System (DAS) for mitigation of anticipated transients without scram (ATWS), and the normal residual heat removal system (RNS) and supporting fluid and ac electrical and systemo for operations during midloop conditions, However, after extensive discussions with the staff, Westinghouse has agreed to expand the number of SSCs covered by RTNSS and to expand the MODES during which R1NSS controls apply, as discussed in the attachment to Westinghouse letter NSD NRC 97 5485, dated December 12,1997. RTNSS oversight is accomplished through administrative controls on the identified SSCs, which specify operability requirements, required actions and the time to accomplish those actions if the operability requirements are not met, surveillance requirements, and the bases for the controls.

However, there are no limiting conditions for operation associated with these RTNSS controls.

The staff has reviewed the administrative controls related to shutdown operations (MODES 5 and 6),

and has identified a concern related to the allowed completion time for required actions during periods of reduced inventory. The proposed administrative control for RNS deing MODES 5 and 6 specify that both RNS pumps should be available prior to entry into MODE 5 with the pressure boundary open or MonE 6 with upper intemals in place and the cavity levelless than full, if one RNS pump subsequently fails, the operator is permitted up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to remove the plant from the MODE in which these controls are applicable. The staff has concluded that this time is excessive when the plant is operating in a reduced 6nventory condition. The short refueling schedules proposed for the AP600 mean the plant could be in reduced inventory conditions for a relatively short time. Thus, it could be possible to enter reduced inventory operations, then have the RNS or one of its supporting SSCs become inoperable, but with the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months acthn completion time, necessary work could be completed and the plant could exit the MODE within the time specified for operator action. (The same action times are specified for RNS support systems, such as component cooling water, service water, and on site ac power.) Thus, for reduced inventory oporations in the applicaole MODES, the 72 hout completion time effi,ctively serves i no safety purpose, The staff thus concludes that the administrative controls on RNS and supporting SSCs for reduced inventory operations dunng MODES 5 and 6 are not conservative and are not consistent with the safe shutdown objective and that action completion times when operability requirements are not met should be more restrictrve and consistent with the length of time the plant is W westinghouse

NRC FSER OPEN ITEM

{

expected to be in reduced inventory operations in the applicable MODES. Resolution of this issue is an Open item.

Responae:

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time to exit the reduced inventory MODE where both RNS CCS and SWS pumps should be available, is appropriate for the following reasons:

- The RNS only provides a nonsafety related means of removing decay heat at reduced inventory conditions. The PXS provides the safety related means to remove decay heat in these conditions through the use of passive feed and blood cooling (RWST injection and ADS venting),

- From a PRA perspective, the PXS capability is more reliable than the RNS / CCS / SWS because of its redundancy, simplicity, and passive nature,

. The PXS features are required to be available by Technical Specification LCOs 3.5.8 (IRWST) and 3.4.14 (ADS). In cases where one of these redundant components is inoperable, the Technical Specifications allow 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore the component. During that time the plant does not have to take action to exit the feduced inventory condition, if the component is not restured within the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Action Time, the plant must immediately initiate actions to remove the plant from the reduced inventory condition. No time limit is specified for leaving the reduced inventory condition.

The plant is only expected to be in reduced inventory for limited times during a refueling outage.

Howevor, in the shutdown PRA the time that the plant was assumed to be in reduced inventory conditions is 120 hr / year. This longer time was selected to cover unexpected delays dunng refueling operations and forced outages for maintenance repairs. In addition, the PRA assumed that th9 reduced inventory condition associated with return to power had the same risk as the one associaled with shutting down. As a result, the PRA over estimates the risk importance of the reduced inventory condition.

The AP600 shutdown PRA takes no credit for the reduced inventory RNS/CCS/SWS short term availability controls. The plant is assumed to stay in reduced inventory for the whole 120 hr / year even when one RNS / CCS / SWS pump fails.

Requinng the plant to make sudden changes in plant operations can lead to increased risk. This philosophy is also used in selecting Technical Specification Action Times.

SSAR Changet None ITAAC Change:

None 440.604F 2 W Westlighouse

IIII N a

NMC FSER OPEN ITEM Question 440.808 (()lTS 6483)

Westinghouse should scrify that all the tech specs related to shutdown operation hase been thoroughly esaluated in concert with the shutdown esaluation report.

Hriponse The techical specifications related to shutdown operation base been thoroughly evaluated in concert with the shutdown evaluation report. Section .i.1 of the SDER presides a sum nary of Technical Specircations related to shutdown modes.

SSAR Resisjont None 0.808N 3 Westingh00$8

7 .

t NRC PSER OPEN ITEM p =m8891 Ovecton 650.25F (OITS 6327)

Re: ,

650.25F: GL 80 009. Low Level Radioactive Waste Disposal his generic letter concerned the requirements for solid waste shipments from a plant. Westinghouse stated that this generic letter is addressed in Section 11.4 of the SSAR, ne staff has revicwed SSAR Section 11.4 and has not found GL 80-09 being addressed. Westinghouse is requested to be specific in addressing this generic letter foi the AP600 design. This is Open item 650.25F.

Derefore, this generic letter is not resolved for the AP600 design.

Response

To the estent that Generic Letter GL.80-009 applies to the design certi0 cation of AP600, it is implicitly addressed in rubsection 11.4.2. To ensure that the COL applicant considers GL.80-009, SSAR subsections 11,4.6 and 11.4.7 will be modified to esplicitly identify the Generic Letter. Note that the SSAR revision shown below also addresses FSER Open item 650.26F.

SSAR Revision:

Subsections 11.4.6 and 11.4.7 revise to state:

11,4.6 Combined Lleense Information for Solid Wnate Management System Procesa Control Program ne Combined License applicant will develop a process control program in compliance with 10 CFR Sections 61.55 and 61.56 for wet sohd wastes and 10 CFR Part 71 and DOT regulations for both wet and dry solid wastes. Process control programs will also be provided by dudors providing mobile or pitable processing or storage systems. It will be the plant operators responsibility to assure that the vendors have appropriate process control progtums for the scope of work being contracted at any partic9l ar time. He process control program will identify the operating procedures for storing or processing wet solid wastes. De mobile systems process 1 control program will include a discussion c" conformance to Regulatory Guide 1.143 (Reference 7), Generic

Lener Gb80 009 (Reference 8) and Generec Lener Gb81039 fReference 9) and, information onf equipment containing wet solid wastes in the nonseismic Radwaste Building.

11.4.7 - References 1-. " Shippers General Requirements for Shipments and Packagings," 49 CFR 173,

2. " Packaging and Transportation of Radioactive Material," 10 CFR 71, .

3. " Domestic Licensing of Production and Utilization Facilities," 10 CFR 50.

650.25F 1

4 NRC FSER OPEN liEM E

4.

Standards fot Protection Against Radiation,* 10 CI'R 20.

5.

Licensing Requirernents for Land Disposal of Radioactise Waste.* 10 Cl~R 61.

6. *USNRC Technical Position on Waste I:ortn/' Rev. I, January 1991, l 7. Regulatory Guide 1.143, " Design Guidancefor Radioactive Waste hianagement Systems, Structures, and Components Installed in Light. Water. Cooled Nuclear Power Plants"

8. USNRC Generk istter Gis30-039, " law 12rel Radioactive Waste Disposal," dated January 29,1980.

l 9. USNRC Generk istter Gl. 81039, "NRC Volume Reduction Polky (Generic litter No. 8139)," dated l Norrmber 30,1981.

650.25F 2 W Westinghouse

NRC FSER OPEN ITEM 0

{p: qi Queston 65026F (OITS 6328)

Re:

650.26F: GL 8139, NRC volume reduction policy This generic letter provided a copy of the Commission policy statement on reduction oflow level radioactive wastes at plant si:ts. Westinghouse stated that this generie letter is addressed in Section 11.4 of the SSAR.

The staff reviewed SSAR Section 11.4 and found that GL Bl.39 was not addressed. Westinghouse is requested to be specific in addressing this generic letter for the AlW10 design. This is Open item 650 26F.

Therefore, this generic letter is not resolved for the AP600 design.

Response

To the entent that Generic Letter GL 81039 applies to the design certification of AF600, it is implicitly addressed in subsection 11.4.2.1. To ensure that the COL applicant considers GL Bl.039, SSAR subsections 11.4.6 and 11.4.7 will be modified to caplicitly identify the Generic letter. Note that the SSAR revision is shown in the response to Open item 650.25F (OITS 6327).

SSAR Reviskn Subsections 11.4 6 and 11.4.7, revise as shown in response to FS11R Open item 650.25 (OITS 6327).

650.26F 1

NRC FSER OPEN ITEM mn -

Ouestion 720.444F (OITS #6182)

The containment layout prevents the formation of diffusion flames that can challenge the integrity of the r.ontainment shell Specifically:

the openings from the accumulator rooms and CVS compartments that can vent hydrogen to the CMT room are eithor located away from the containment wall and electrical penstration junction bores, or are covered by & secure hatch, and IRVJST vents near the containment wall are oriented to direct releases away from the containment shell.

These provisions need to be confirmed by ITAAC WEC has not provided this ITAAC. This is Open item 720.444F.

Response

Diffusion flames at the openings from the CVS room are not a significant challenge; they occur for- 1%

of the core melts. As a result, it is not necessary to provide an ITAAC for the openings from this room.

The frcavency that diffusion flames may occur at the IRWST vents is also low (- 5% of the core melts).

In addition, the NRC staff has not agreed with Westinghouse that the design of the IRWST vents is capable of preventing diffusion flames from challenging the containment integrity. Therefore, it should not be necessary to provide an ITAAC for the IRWST vent orientation.

The frequency that diffusion flames may occur at the openings from the PXS valve / accumulator rooms is twre significant (- 26% of the core melts). The location of these openings will be subject to an ITAAC.

SSAR Change:

An attachment is provided which addJ a discussion of the openings from the PXS valve / accumulator rooms to SSAR subsection 6.2.4.5.

ITAAC Change:

An design commitment and an inspection is added to the containment hydrogen control system ITAAC (2.3.9) as shown in the attachment. The inspection requires a minimum distance between the openings trom the accumulator rooms and the containment shell.

g gg, 720.444F 1 L

NRC FSER OPEN ITEM

8tammtr Revision to SSAR subsection 6 2.4.5 6.2.4.5 Tests and Inspections 6.2.4.5.1 Preoperational inspection and Testing Hydrogen Monitoring Subsystem Pre-operational testing is performed either before of after installation but poor to plant startup to venfy performance.

Hydrogr5 Recombination Subsystem Pro operational testing is performed following vendor production testing and installation but prior to plant startup to venty PAR performance. The PAR's are verified to provide a hydrogen depletion rate of greater than or equal to the minimum depletion rate identified in Table 6.2.4 2. It is also verified that two PAR's are installed within containment at an elevation of between 150 and 175 feet with the PAR centerline at least 10 foot from the containment shell, it is also verified that a PAR is located in the exhaust of an IRWST vent and within the chemical and volume control system compartment.

A sample of the PAR cartridges or plates are selected and removed from each passive autocatalytic recombiner and surveillance bench tests are performed on the removed specimens to confirm continued satisfactory performance. The specimen is placed in a performance test apparatus and exposed to a known standard air / hydrogen sample. The test instrumentation will be designed to assess PAR performance and the time to reach a threshold recombination start to measure degradation in catalytic action. The overall PAR performance verification will be based on vendor testing recommendations and may include among other means, recombiner intemal or exhaust temperature measurement or exhaust sample concentration measurement. Should intemal temperature measurement be utilized as the measured recombination parameter, location of the sensor must be consistent for all samples and with vendor test recommendations to assure consistency between tests, The recombiner start vidification will be based on a time dependent measurement of the recombination rate parameter or other instrumentation venfying the recombination start. The vendor manufacturing acceptance data or accepted industry standards will be utilized as acceptance data provided it represents portormance in excess of the required rate specified in Table 6.2.4 2.

Hydrogen Ignition Subsystem Pre. operational testing and inspection is performed after installation of the hydrogou ignition system and prior to piant startup to venfy operability of the hydrogen ignitors it is verified that 64 igniter assemblies are installed at the locatioits defined by Figures 6.2.4 5 through 6.2.411, Operability of the igniters is contirmed by verification of the surface temperature in excess of the value specified in 72 - 2 W wesunnon.

l NRC FSER OPEN ITEM i

W Tab!e 6 2 4 3. This temperature is sufficient to ensure ignition of hydrogen concentrations above l

the flammability limit.

Pre-operational inspection is performed to verify the location of openings through the ceilings of the passive core cooling system valve / accumulator rocms. The primary openings must be at least 19 feet from the containment shell Primary openings are those tht,t constitute 98% of the opening area. Other openings must be at least 3 feet from the containment shell.

720.444F 3

[ W85tingh0050

NRC FSER OPEN ITEM met {

Rension to ITAAC Table 73 9 S Table 2.3.9 3 (Cont.)

Insptctions. Tests. Analyses, and Acceptance Criteria lhesign Commitment inspections Tests. Analyses Acceptance Criteria 4.b) The VLS provides PAR I) Inspection for the cAlstence 1) Two PAR devices are provided inside devices for control of the of two PAR devices inside containment within the upper compartment containment hydrogen containment will be performed, between elevations 150 and 175 ft, concentration during and following a design basis 11) Type tests, analyses, or a 11) A report exists and concludes that the accident. combination of type tests and PAR depletion rate for each installed PAR analyses will te performed on is greater than or equal to I scfm of the PARS. hydrogen at a prevailing concentration of 3 volume percent for a test conducted at atmospheric pressure +2 psi and an ambient temperature of 120.

5.1he VLS provides the 1) Inspection for the numtwr of i) At least M hydrogen igniters are nons:dety related it.netton to igniters will be perfonned, provided inside containment at the control the containment locations specified in Table 2.3.9-2.

hydrogen concentration for beyond design basis accidents, li) Operability testing will be 11) The surface temperature of the igniter performed on the igniters. exceeds 1700*F.

iii) An inspection of the as- lii) The minimum distance beiween the built containment internal primary openings through the ceilings of structures will be performed. the passise core cooling synem valse /

accumulator rooms and the containment shell is at 'tast 19 feet. Primary openings are those that constitute 9804 l of the opening area. Other openings through the ceilings of the:2 rooms must be at least 3 feet from the containment shell, 720.444F 4 g ,

NRC F8ER OPEN ITEM 1

Question: 720.451 F (OITS #6387) re: Generic Fragility Data (frorn SER input Section $5.2.5)

Generic fragility data were used when insufficient infortnation was available to define the liCLPF value using one of the methods described above. Rose cases where this approach was used were:

Reactor internals and core assembly that includes fuel

Control rod drive mechanism (CRDM) and hydraulic drive units

Reactor coolant purnp

Accumulator tank

P: ping a Cable trays

Valves

Itaitery tacks

Main control room operathn and switch stations

Ceramic insulators Reference 10 was used for all of the components listed above except ceramic insulators, for which recognized industry low fragility data were available, ne staff believes that the suggested generic fragility values are intended for a preliminary analysis only. Dese generic values should not be used for critical components which are important to plant risks. In addition, for components with new design features, it should be confirmed that the new design features do not potentially contribute to lowering fragility values. Such examples may include the fuel rods, for which some diffeiences in design (e g.. different outside diameter and additional gas space below the fuel pellets) .ae observed compared with the typical four loop design.

Westinghouse responded that Reference 10 provides a summary of generic fragility data for preliminary analysis only; howeser, they are rep?csentative of the anticipated capacity. Westinghouse has identified a COL item which requires verification of as built conditions conforming to the seismic margin evaluation. It is stated in the AP600 PRA Chapter 59 Section 59.10.6, "De combined license applicant referencing the AF600 certified design will confirm that the as built plants conforms to the design used as the bases for the seismic margin evaluation"(RAI 230.136).

The serification of as. built conditions is discussed in Subsection $$.3 of this FSER.

The staff requested information on llCLPF margin for rigid components with non-ductile supports. De SSE design load and the RLE for the AP600 are 0.3g and 0.5g, respectively. Derefore, a llCLPF margin of 1.67 is implied for all the safety related equipment and components. T achieve this HCLPF margin, a inedian margin factor of at least 4.2 is needed. This is based on assumptions that a relatively low variability of Bc of 0.40 is used for a fragility estimate. and the seismic design is performed up to the limits of the code design allowables.

For relatively flesible/ ductile components, such as piping, the design criteria in the PRA is considered to gi'e 4 sufficient margin to achiese the above median factor of 4.2. Howeser, for dynamically rigid components whose support structures are considered to have a non-ductile failure mode, such as clastic buckling and shear failure .'n 720.451F 1

m. _

NRC FSER OPEN ITEM 1 ym iy!

l fillet welds or anchor bolt joints, the design requirements in the SSAR may not be sufficient to provide this safety rnargin. According to the staffs estimate, an additional median margm factor of 2.1 to 3.0 is necessary to ashieve the aforementioned llCLPF margin of 1.67 for relatively rigid components with non-ductile suppon structures.

Westinghouse responded that the components in the AP600 design generally have margin factore in excess of me rar of 2.1 to 3.0. De calculated margins of several specific examples were included in the response w RAI 230.D9, Revision 1.

For the hypothetical case where the component has rigid response characteristics with non ductile wpport structures.

Westinghouse stated that gt:ncric fragility data that is in the public domain, and is also used in Reference 10, does not reflect components having low IICLPF values. Derefore, this hypothetical case would be very plant specific.

De AP600 component support designs do not deviate from those seen in other plants and reDected in the generic fragil ty data, and, therefore, have HCLPF salues below seismic margin requirements. If this hypothetical case did esist, it would be a plant specific case, and it would probably not provide the seismic margin commitments of AP600, unless the as built plant is verified to conform to the seismic margin of 0.5g. It is stated in PRA Chapter 59, Section 59.10.6, "De combined license applicant referencing the AP600 certified design will confirm that the as built plants conform to the design used as the bases for the seismic margin evaluation." Derefore, this case would be identified and addressed by the COL applicant (RAI 230.139, R D.

However, it is not clear how an as built verification program would identify such deficiencies in median margin factors for other supports which may be designed up to the code allowable values. His is required to be serified by a COL Action item. His is an Open item.

Responce:

His open item states that a COL action item is required because it is not clear hev. an as built venfication program would identify such deficiencies in median margin factors for othei supports which may be designed up to the code allowable values.

The COL action itera reported in AP600 PRA Chapter 59 will be revised as shown below.

PRA Revision:

PRA subsection 59.10.6 will be revised as follows:

De Combined License applicant referencing the AP600 certified design shouldperform a seismic walkdown to wW confirm that the as built plant conforms to the design used as the basis for the seismic margin evaluation and to assure that seismic spatial systems interactions do not exht. Details of the seh,nic walkdown will be developed by th< Combined License applicant.

The Combined License apeticant referencing the AP600 certyled design should compare the as built SSC HCLPFs to those assumed in the AP600 sekmic margin evaluation. Deviationsfrom the HCLPF values or

) assumptions in the schmic margin evaluation should be evaluated by the Combined License applicant to f determine if unacceptable vulnerabilities have been introduced.

720.451F 2

NRC FSER OPEN ITETA -

1p 4 Question: 720A52F . (OITS #6388) re: Verification of Equipment Fragility Data (from SER input Section 55 3)

Since no walldowns can be performed at this time, the staff requested Westinghouse to show how'the key assumptions for structures, systems, and components considered in the SMn can be veri'ied for the as-built anf as-operated plant conditions. Examples of this include proper anchorage of equipment and seini fragility of electrical / electronic equipment which may be different in the future. Westinghouse responded th,'s : !ication that

- the as built plant confirms the basis of the seismic margin evaluation will be performed by the CA c icant (RAI 230.115).

' Westinghouse stated in Section $$.2.2.5 of AP600 PRA that the seismic n.argin evaluation has focused on demonstrating that the design of nuclear island structures, safety related equipment, and equipment supports can carry the loads induced by the RLE. His evaluation incorporates as-specified equipment data. After the plant has been built, it will be necessary to perform a verification of the seismic margin assessment for the installed conditions.

De AP600 PRA Section 59.10.6 Revision 9 provides the COL information for the AP600 PRA, including the SMA.

De COL action item for seismic margin, as stated in the PRA, is t!- COL applicant referencing the AP600 certified design will confirm that the as built plant conforms to the design used as the basis for the seismic margin evaluation.

It is the responsibility of the COL applicant to define how this confirmation is performed (RAI 230.112).

De staff agrees that there needs to be a verification program to confirm the data and assumptions made in the SMA for all items. De description of how this will be accomplished is lacking and should be described. He process of identifying what data and assumptions need to be verified, how and where they will be documented, and how the verification process will be conducted by the COL spplicant should be included in the AP600 PRA. For example, where generic fragility data from Reference 10 was used, will a new SMA be performed by the COL applicant to confirm the assumed HCLPF values? Based on the above discussion, this is an Open Item.

Response

This open item states that a COL action item should be added to the PRA that explains the certification program to confirm the data and assumptions made in the SMA. e The COL action item reported in AP600 PRA Chapter 59 will be revised as shown below. Note it is the responsibility of the Combined License applicant to develop the specific details noted by the NRC above.

PRA Revision:

PRA subsection 59.10.6 will be revised as follows:

'! ' The Combined License applicant referencing the AP600 certified design shouldperform a seismic walkdown i~ te will confirm that the as built pl9nt conforms to the design used as the basis for the seismic margin evaluation

[

1 asw assure that seismic rpaties systems interactions dc not exist. Details of the seismic walkdown will be i developed by the Combined License applicant.

720.452F-1

NMC FSER OPEN ITEM

!!i; "M l

l The Combined License applicant referencing the AP600 certifled design should compare the as built SSC .

l HCLPFs to those assumed in the AP600 seismic margin evaluation. Deviationsfrom the HCLPF values or l assumptions in the seismic margin evaluation should be evaluated by the Combined Ikense applicant to l determine (f unaccepwble vulnerabilities have been introduced.

720.452F-2 gg

NRC FSER OPEN ITEM Question: 720.453F (OITS #6389) re: Spatial Interaction (from SER inp- Section 55.4)

The staff requested Westinghouse that spatial interactions (e.g., seismic impact between adjacent components and Seismic 11/1 interactions) be included in the SMA. Westinghouse responded that the interaction between the Turbine Building and the north end of the Auxiliary Building is explicitly discussed in PRA Section 55.5.8 (RAI 230.ll4).As part of SMA, the seismic interaction between the Turbine Building and the Nuclear Island was evaluated. It was determined that:

The adjacent Auxiliary Building structural integrity will not be lost with the failure of the Turbine Building.

e it is not likely that the size and energy of debris from the Turbine Building will be large enough to result in penetration through the Auxiuary Building roof structure.

Even though it is not likely that the Turbine Building debrb could be large enough or have sufficient energy for penetration through the Auxiliary Building roof structure, this event was evaluated. He consequences of damage to the safety related equipment in the Auxiliary Building was investigated. It was determined from this investigation that, should an event occur that causes the failure of equipment in the upper elevations of the Auxiliary Building, it.e results of the SMA analysis, the plant HCLPF value, and the insights derived from the SMA would not be affected. Moreover, according to the AP600 focused PRA realts, steam line break events that would result from damage to equipment in upper elevations are not dominant contributors to the core damage frequency. Further, any lots of equipment in the upper elevations would not affect the passive safety systems used to put the plant in a safe shutdown condition should an event occur.

The information presented in Sections 55.5.8 and 55.2.2.6 of the AP600 PRA does address the concern of seismic interaction between the Turbine Building and the Auxiliary Building. In the AP600 PRA Section 55.3.3, Annex Building. Diesel Generator Building, and Radwaste Building are assumed to have failed for the SMA. No credit is taken for systems in those buildings. De interaction between the other building and the Nuclear Island is assumed to have no detrimental effect on the Nuclear Island structures. However, the AP600 PRA does not address how the failure of the Annes Building and/or the Radwaste Building affects the safety related structures and components of the Nuclear Island. De concern of interaction effects also includes potential impact from deflection of adjacent components or collapse of non. seismic Category I structures and components. His is required to be verified by a COL Action item. his is an Open item.

Response

De NRC has stated that the AP600 PRA does not address how the failure of the Annen Building and/or the Radwaste Building affects the safety related structures and components of the Nuclear Island. He concern of interaction effects also includes potential impact from deflection of adjacent components or collapse of non-seismic Category I structures and components. His is required to be verified by a COL Action item.

It is noted that r.o action is required related to the Annex Building and/or the Radwaste Building for the following reasons:

T west!nghouse

NRC FSER OPEN ITEM The Annex Iluilding is Seismic Category 11 which in essence assures that similar margin as those associated with Seismic Category I structures (i.e., ilCLPF 2.0.5g), and therefore, this building need not be considered further with respect to spatial interaction.

He Radwaste !!uilding failure on the nuclear island structures has t>een evaluated by Westinghouse, and this evaluation has been reviewed and found acceptable by the NRC and t.ute is nn need to consider spatial interaction associated with this building further. This evaluation is described by the NRC in their Open item 220.ll6F (OITS

6303). The NRC has stated:

"In Revision 12 of SSAR Section 3.7.2.8, Westinghouse, based on the energy balance theory, provided the analysis procedures which are to be used for demonstrating that the collapse of the radwaste building will not cause any damage of the NI structures. Itecause the application of energy balance for checking potential damages of structures is consistent with the industry practice, it is acceptable to the staff. Also, in the August iI through 15,1997 meeting, the staff reviewed the final calculation (Calculation No. 5000-S2C-002) and found that the analysis procedure described in the SSAR was properly applied in the evaluation of the impact between the N1 and the radwaste building and that the impact from the radwaste beilding in the event of an SSE would not impair the integrity of the N1. His is acceptable to the staff regarding the potentialinteraction between the radwaste building and the Ni structures. ."

The COL action item reported in AP600 PRA Chapter 59 will be revised to include a statement about spatial interaction as shown below.

PRA Revision:

PRA subsection 59.10.6 will be revised as follows:

! The Combined License applicant referencing the AP600 certified design should perform a seismic walkdown

! to wdl confirm that the as-built plant conforms to the design used as the basis for the seismic margin evaluation 1 and to assure that seismic spatial systems interactions do not exist. Details of the seismic walkdown will be developed by the Combined License applicant.

The Combined License applicant referencing the AP600 certified design should compare the as built SSC llCLPFs to those assumed in the AP600 seismic margir, evaluation. Deviationsfrom the HCLPF values or assumptions in the seismic margin evaluation should be evaluated by the Combined License applicant to i determine {f unacceptable vulnerabilities have been introduced.

720.453F-2 W- WestInfiouse to Westinghouse Letter DCP/NRCl209

.lanuary 9,199d l

l l

FSER Open hem m By Open Item 220.128F (OITS # 6315)

Because of the complication of the coupled shield / auxiliary building structures, Westinghouse informed the staff that the completed stmetural design of this building will not be performed. (The five-story auxiliary building is stmeturally connected with the cylindrical shell shield building at six different elevations and formed a coupled structure. The coupled shield / auxiliary building is founded, together with the containment vessel and the containment internal structures, on a irregular shaped foundation mat.) Instead, the detailed design would be completed only for the critical sections of stmetures. As described in revision 12 of SSAR Section 3.8.4.5.3, Westinghouse identified 9 critical sections for which Westinghouse completed its stmetural design. The staff reviewed samples of these critical section designs and raised the following concems:

+ In reviewing the design calculations for the auxiliary building roof slab at Elevation 180 ft (Calculation Nos.1260-SSC-003, Revision 2, and 1260-CCC-003, Revision 3), the staff identified two issues:

(1) The design did not account for the effect of global out-of-plane seismic moments along the edge of the roof slab.

(2) Reinforcements for the concrete slab in the north south direction (parallel to floor steel girders) along the roof edge should be designed assuming no composite action of the concrete slab w% the steel girder.

. The design of the shield building roof structures is not adequate as discussed under Open item 3.8.4.4 2 above.

. Westinghouse should include the detailed design drawing for each of these critical sections in the SSAR.

Westinghouse needs to revise the design calculation to addrest. the staff's concern discussed above and provide figures describing reinforcement details of critical sections in the SSAR.

Response

The issues related to the design of the auxiliary building mof were addressed in Westinghouse's letter of November 19,1997, DCP/NkCl143 Westinghouse had proposed that the critical sections details be provided in a summary report to be referenced from the SSAR. As stated in FSER Open item 220.122F, a draft design summary report (Design summary report - Auxiliary Building Structures,1200-S3R-001, revision 0 (draft) was available for review during the meeting on January 14 through 16,1997. In a meeting open item (OITS # 5150) from this meeting (summarized in the March 18,1997 letter, Attachment 3), NRC staff identified that Westinghouse should include critical section details in a formal revision of the SSAR.

Westinghouse has now prepared information on the critical sections in the auxiliary and shield buildings for inclusion in the SSAR in a new Appendix 3H. This new appendix is substantially the sane information as was in the draft summary report already reviewed by the staff. The draft design summetry reports that were reviewed by the staff in previous meetings are being issued as internal W Westiaghouse 220.128-1

= Psan open nom 1 .

. Westinghouse documents. These design summary reports will be reconciled for as procured or as-built conditions as described in SSAR subsection 3.8.4.5, SSAR Revisiont _

Revise last two sentences of second paragraph of subsection 3.8.4.4.1 to reference typical details in Appendix 3H asfollowsi Rge: 3.8.4 2 :. hew; typia! dat w:!!: =d-t' - *aangement& m afercing ;te ! Fign= 3.8.4 3 I :. hews typ!=! ahfc=ing fer 6: :.!:b:. Appet 4 311 describes the design of typical shear walls and I floor slabs.

Revise fish paragraph of subsection'3.8.4.4.1 to reference typical details in Appendix 3H as follows: ,

The structural steel framing is used primarily to suppon the concrete slabs and roofs. Metal decking, supported by the steel framing, is used as form work for the concrete slabs and roofs. He structural I steel framing is designed for vertical loads. Mgn= '..S.4-4-Appendix 311 shows typical structural steel framing in the auxiliary building.

Revisefirst paragraph of subsection 3.8.4.5 asfollows:

A design summary repon b prepared for seismic Category I structures documenting that the structures meet the acceptance criteria specified in subsection 3.8.4.5. Ref=== 50 previ&r. the I &rJg :. =.j = pert - Ca!=! r.=t!=; incia&d in S: = pert :=: ne design of representative I critical elements of the following structures is described in Appendix 311.

Passive containment cooling system water storage tank

- Shield building roof to cylinder connection

Shield building to auxiliary building connection at elevation 180'

- South wal: of auxiliary building (column line 1)

Interior wall of auxiliary building (column line 7.3)

West wall of main control ;vom in auxiliary _ building (column line L), elevation 117'-6" to I elevation 153'-0"

North wall of auxiliary building (column line 11 between Q and P), elevation ll?'-6" to elevation 153'-0"

Floor slab in north end of auxiliary building at elevation 135'-3" including:

- 9 inch concirte slab on metal deck

- - 24 inch reinforced concrete slab

- 24 inch finned floor above the main control room 1- -* Spent fuel pool dM&r wn!! =d ficarstructural module Revise Figure 3.8.4-6 (Sheet 1 of 3) as shown in attachment Add new Appendix 3H as shown in attachment

Delete Figures 3.8.4 2, 3.8.4 3 and 3.8.4-4 MM 220.128-2

FSER Open item' my Other SSAR Revisions and ITAAC Revisions The information included in Appendix 3H reflects design calculations that were updated to include

.changes n idi es gn cr erita io tre o vs l e NRC staff comments during the structural meetings. Some of the information in Appendix 3H is inconsistent with information currently in the SSAR. SSAR Table 3,8,4 7 and Certified Design Table 3.3.1.will te updated to reflect the design calculation changes and to be consistent with the information in Appendix 3H in SSAR Rev 20. The changes to Table 3.8,4 7

. and CDM Table 3.3.1 will be changes to the information in the columns that specify the required and.

the provided amount of reinforcement. 'lhese changes are not required to evaluate the ret.ponse to this open item.

o EMN 220.128-3

m Appendix 311 ACxiliary llullding CriticCl Sections 311,1 Introduction Tids appendix summarizes the structural design and analysis of structr identified as

" Critical Sections" in the auxiliary and shield buildings. The design summiuu include the following information:

Description of buildings

Goveming codes and regulations

Structural loads and load combinations

Global analyses

Simctural design of critical stmetural elements Subsections 311.2 through 3H 4 include a general description of the auxiliary building, a summary of the design criteria and the global analyses. The twelve critical sections are identified in subsection 311.5 and shown in Figures 311.5-1 (3 sheets). Representative design details are provided for these structures in subsection 311.5.

311.2 Description of Auxiliary lluilding The auxiliary and shield buildings are reinforced concrete structures. The auxiliary building is one of the three buildings that make up the nuclear island and shares a common basemat with the containment building and the shield building. The auxiliary building general layout is shown in Figure 311.21. It is a C-shaped sectio" of the nuclear island that wraps amund appmximately half of the circumference of the shield building. The building dimensions are shown on key structural dimension drawings, Figure 3.7.2-12.

The auxiliary building is divided into six areas, which are identified in Figure 3H.21. It is a 5-story building; three stories are kicated atxwe grade and two are located below grade.

Areas I and 2 (Figure 311.2-1) have five floors, including two floors below grade level. The lowest floor at elevation 66'-6" is used exclusively for housing battery racks. 'Ihe next higher floor, at elevation 82'-6", also has battery racks and some electrical equipment. The floor at the grade level, elevation 100'-0", has electrical penetration areas, a remote shutdown workstation room, and some Division A and Division C equipment. The main contml mom is situated on the floor at elevation i17' 6", which also has moms for the main steam and feedwater lines. The floor at elevation 135'-3" carries air filtration and air handling units, chiller pumps, and other mechanical and electrical equipment. The roof for areas I and 2 is at elevation 153'-0".

Areas 3 and 4 of the auxiliary building are the areas cast of the containment shield building.

Valve and piping areas, and some mechanical equipment, are located in the basement floor at elevation 66'-6" The floor at elevation 82'-6" has a piping penetration area, a radiation chemistry laboratory, makeup pumps, and other mechanical equipment. The floor at grade level elevation 100'-0" has an electrical penetration room, a staging area for the equipment hatch, and the access opening to the annex building. The electrical penetration area, trip switchgeam, and motor control centers occupy most of the floor at elevation i17'-6". The Revision: 20 (Draft)

T Westingh00S8 311-1 January 9,1998

Appendix 311 Armitiary fluilding Critical Sections fkor at elevation 135' 3" is used for the storage of main control room air cylinders and pnivides access to the annex bu!! ding. The roof for these areas is at elevation 160'-6".

Areas 5 and 6 include facilities for rtorage and handung of new and spent fuel. The spent fuel pool, fuel transfer canal, and cask loadmg/washdown pit have concrete walls mal floors.

They are lined on the inside surface with stainless steel plate for leak prevention. The walls and major floors are constructed using concrete filled steel plate modules. The new fuel storage area is a separate reinforced concrete pit providing tempora y dry storage for the new fuel assemblies. A 150-ton cask harxiling crane travels in the east-west direction. The location and travel of this crane prevents the crane from carrying loads over the spent fuel pool to preclude them frorn falling into the spent fuel pool. Mechanical equipment is also kicated in this area for spent fuel cooling, residual heat removal, and liquid waste processing.

This equipment is generally nonsafety-related.

The shield building fonus 4ma 7 of the auxiliary building. This appendix describes criucal sections in the shleid bcilding roof and its connection to the cylindrical wall.

311.3 Design Criteria lhe auxiliary and shic!d building stmetun s are reinio.ced concrete structures, structural modules, and horizontal concrete slabs supported by composite structural steel framing.

Seismic forces are obtained from the response spectrum analysis of the three-dimensional finite element analysis models. The shear wall and floor slah design also considers out-of-plane bending and shear forces due to loading, such as live load, dead load, scismic, lateral carth pressun , hydrostatic, hydnx!ynamic, mul wind ressure.

The shield building roof and the passive contairunent cooling water storage tank are analy7ed using three-dimensional finite element mcxtels with the ANSYS and GTS111UDL computer codes. Loads and load combinations include construction, dead, live, thennal, wind, aint seismic. Seismic loads are applied t. equivalent static accelerations. The seismic msponse of the water in the tank is analy7ed in a separate finite element response spectrum analysis with seismic input defined by the Ikor response spectrum.

The stmetural steel framing is used primarily to support the concrete slabs and nofs.

Metal decking, supported by the steel framing, is used as fonn work for the concrete slabs aral roofs.

Revision: 20 (Draft)

W Westinghouse 311-2 January 9,1998

Appendit 311 A;xiliary llullding; Critical Sections

The finned floors for the main control mom and the instrumentation and contml room ceilings are designed as reinforced concrete slabs in accordance with American Concrete Institute standard ACI 349. The steel pancis are designed and constmeted in accordance with American institute of Steel Construction Standard AISC N690. For positive bending, the steel plate is in tension and the steel plate with fin stiffeners serves as the bottom reinforcement. For negative bending, compression is resisted by the stiffened plate and tension by top reinforcement la the concrete.

311J.1 Governing Codes and Standards The primary codes and standants used in the design of the auxiliary and shield buildings are listed below:

ACI 349-90, " Code Requirement for Nuclear Safety-Related Structure Steel" (refer to subsection 3.8.4.5 for supplementary requirements)

ACI 318-95 " Building Code Requirements for Reinforced Conmte" (refer to subsection 3.8.4.4.1 for applicability).

ANSI / AISC N690-1984, " Specification ter the Design, Fabrication and Eirction of Safety-Related Steel Structures for Nuclear Ivilities" (refer to subsection 3.8.4.5 for supplemental requirements).

311.3.2 Seismic Input The SSE design response spectra are given in Figures 3.7.1-1 and 3.7.1-2. They are based on the Regulatory Guide 1.60 response spectra anchored to 0.30g, but are amplified at 25 liertz to 'ellect larger high-frequency seismic energy content observed for castem United States sites. The seismic analyses consider a range of soil conditions with shear wave vehicity greater than 1000 feet per second. The nuclear island seismic analyses are summarized in section 3.7.2.

311.3.3 Loads The auxiliary and shield buildings are seismic Category I structures. The loads listed in the following subsections are used for the design of the building stmetures. All the listed loads are not necessarily applicable to all structures and their elements. Loads for which each structural element is decigned are based on the conditions to which that particular structural element is potentially subjected.

Dead Load (D):

The weight of all pennanent construction and installations, including fixed equipment, is included as the dead load during its normal operating condition.

Rev.sion: 20 (Draft)

[ W851111gh0US8 311-3 January 9,1998

Appendis 311 A~siliary ll:llding Critic 1 Sections The weight of minor equipment (not specifically included in the dead load), piping, cables and cable trays, ducts, and their supports was included as equivalent dead load (EDL). A minimum of 50 pounds per square foot (psf) was used as EDL, For floors with a significant number of small pieces of equipment, the total weight of miscellaneous small pieces of equipment, divided by the floor area of the room plus an acklitional 50 psf was used as the equivalent dead load.

Earth Pressure (II):

1he static carth pressure acting on the structures during nonnal operation is considerea in the design of exterior walls. The_ dynamic soil pressure, induced during a safe shutdown canhquake (SSE), is included as r. scismic load.

I.ive Loads (L):

The load imposed by the use and occupancy of the building is included as the live load. Live loads include floor ama toads, laydown loads, fuel transfer casks, equipment handling loads, trucks, railroad vehicles, and similar items. The ikx)r area live load is not applied on amas occupied by equipment whose weight is specifically included in the dead load. Live load is applicable on areas under equipment where access is pmvided, for instance, the floor under an elevated tank supported on legs.

Floor loading diagrams are prepared for areas for component laydown. 'lhe diagrams show the hication of major pieces of equipment and their foot-print loads or equivalent unifomily distributed loads.

The following live load items are considered in design:

A. Building floor loads The following minimum values for live loads are used, Stmetural platfonns and gratings 100 psf Ground ikxirs 250 psf All other elevated flocits 200 psf (This load is reduced if the equivalent dead load for the floor is mom than 50 psf. The sum of the live load and the equivalent dead load is 250 psf.)

Revision: 20 (Draft)

[ Westifighouse 311 4 January 9,1998

== -

~

Appendis 3H Ausiliary fluilding Critical Sections l

' B. Roof lomis - -

%c roof is designed for a unifomi snow load of'63 psf calculated in accordance with =

ASCE 7 88. This corresponds to ground snow load of 75 psf exposure factor of 1.0, thennat factor of 1.0, and an importance factor of 1.2.-

C. Concentrated lomis for the design of local members Concentr.*ed load on beams and 5,000 pounds so applied as to maximize girders (in load combinations that moment or shear, his load is not

- do not inchide seismic load) carried to columns or walls. It is not -

applied in amas where no heavy equipment will be located or transported, such as the access control areas.

Concentrated load on slabs 5,000 pounds so applied as to maximize -

(considered with dead load only): moment or shear. His load _is not carried to columns or walls, it is not applied in access contml areas.

In design reconciliation analysis, if actual loads are established to be lower than the

- above loads, the actual loads are used for :econciliation.

D. Temporary exterior wall surcharge When applicable, a minimum surcharge outside and adjacent to subsurface wall of 250 psf is applied.

E. Construction h) ads he additional construction loads imiuced by cranes, trucks, and the like, with their ,

pickup loads, are considered. For steel beams supporting concrete floors, the weight of the wet concrete plus 100 psf unifonn load and 5,000 pounds concentrated load, distributed near points of maximum shear and moment, is applied. A one-thiniincrease in alhiwable stress is pennitted. ,

Metal decking and precast concmte panels, used as formwork for concrete floors are designed for the wet weight of the concrete plus a construction lise load of 20 psf-uniform or 150 pounds concentrated. De deflection during normal operation is limited to span in inches divided by 180, or 0.75 inch, whichever is less.

F. Crane lomis ne impact allowance for traveling crane supports and runway horizontal forces is in accortlance with AISC N690.

Revision: 20 (Draft)

T Westinghouse 311-5 January 9,1998

Appendix 31I Atxiliary Itilding Critic-l Sections O. Elevator loads Die impact allowance used for the elevator supports is 100 percent, applied to design capacity and weight of car plus appurtenances, unless otherwise specified by the equipment supplier.

11. Equipment laydown and major maintenance Floors are designed for planned refueling and maintenance activities as defined on equipment laydown drawings.

Wind Load Vic wind loads are as follows:

Design wind (W)

For the design of the exterior walls, wind loads are applied in acconlance with ASCE 7-88 with a basic wind speed of 110 mph. 'The importance factor is 1.11, and the exposure category is D. Wind loads are not combined with seismic loads.

Tomado load (W,)

The exterior walls of the auxiliary and shield buildings are designed for tomado. A maximum wind speed of 300 mph (maximum rotational speed: 240 mph, maximum translational speed: 60 mph) is used to design the structures.

Seismic Loads (E,)

The SSE (E,) is used for evaluation of the structures of the auxiliary and shield buildings.

E, is defined as the loads generated by the SSE specified for the plant, including the associated hydnxtynamic loads and dynamic incremental soil pressure (based on two-dimensional soil stmeture interaction [SSl] analysis results).

Operating Thermal 1,oads (T,)

Nonnat thennal loads for the exterior walls and roofs are addressed in the design. These correspond to positive and negative linear temperature gradients with the inside surface at an average 70 F and the outside surface at -10 F and +100 F, respectively. These loads are considered for the seismic Category I structurer in combination with the SSE also. All exterior walls of the nuclear island above grade are designed for these thermal loads even if the exterior surface is protected by an adjacent building. The thennal gradient is also applied to the portion of the shiehl building between the upper annulus and the auxiliary building.

Revision: 2n (Drafl)

W Westinghouse 3H-6 January 9,1998

Appendix 311 Artillary llullding Critical Sections Normal thennal loads for the passive containment cooling system (PCS) tank design are calculated based on the outside air temperature extremes specified for the safety-related design. With the water temperature in the tank assumed at +40'F, the positive and negative temperature gradients are detennined for the outside surface at -40*F and +115*F respectively.

P.onnal thennal loads due to a thermal gradient in the stmetures below the grade level (uterior walls and basemat) an: small and are not considered in the design.

F.ffects of Pipe Rupture (Y)

The evaluations consider the following Ic As:

Accident design pressure load, P., within or across a compartment and/or building generated by the postulated pipe rupture, including the dynamic effects due to the pressure time history.

Main steam isolation valve (MSIV) and steam generator blowdown valve companments are designed for a pressurization load of 5 pounds per square inch (psi).

Accident thennal loads, T , due to thermal conditions generated by the postulated pipe break arul including T .

Temperature gradients are based on an exterior air temperature of -10 F.

Die stmetund integrity of the west wall of the main control room is also evaluated for the jet impingement (Y,)

311.3.4 Load Combinations and Acceptance Criteria Concrete structures are designed in acconlance with ACI 349 for the load combinations and load factors given in Table 3.8.4-2. Steel structures are designed in accordance with AISC N690 for the load combinations and stress limit coefficients given in Table 3.8.4-1. The following supplemental requirements are applied for the use of AISC N690:

In Section Ql.0.2, the definition of secondary stress applies to . stresses developed by temperature loading only.

In Section Ql.3, where the structural effects of differential settlement are present, they are inchaled with the dead load, D.

in Table Ql.5.7.1, the stress limit coefficients for compression are as follows:

1.3 instead of 1.5 in load combinations 2,5, and 6 1.4 instead of 1.6 in load combinations ~'. 8, and 9 1.6 instead of 1.7 in load combination i1 Revision: 20 (Draft)

T Westinghouse 311-7 January 9,1998

Appendix 311 A xiliary llullding Critled Sections

in Section Ql.5.8, for constrained a mbers (rotation and/or displacement constraint such that a thennalload causes significant stresses) supporting safety-related structures, systems, or components, the stresses under load combinations 9,10, and i1 are limited to those allowed in Table Ql.5.7.1 as modified above.

311.4 GLOllAL SEISMIC ANALYSES A global seismic analysis of the AP600 nuclear island structure is perfonned to obtain building seismic response spectra for the seismic design of nuclear safety-related structures.

This analysis is described in subsection 3.7.2. For detennining the out-of-plane seismic loads on slabs and wTil segments, spectral accelerations are obtained from the relevant response spectra, using the 7 percent damping curve, lland calculations are performed to estimate the out-of plane seismic fonrs and the corresporxling bending moment in each shear wall and floor slab element to supplement the loads obtained from the response spectra analyses.

The in-plane r.cismic loads for the design of the shear walls and the slabs in the auxiliary building are based on a response spectrum analysis of the auxiliary building and the shield building 3D finite element models. The response spectrum analyses are perfomled for two cases: one that considers the reinforced concrete elements to be uncracked with full clastic stiffness, and the other that models the elements with 70 percent of their full stitmess. The larger of the two values for each finite element, from these two cases, for the stress resultants is used in the design evaluation.

311.5 STRUCTURAL DESIGN OF CRITICAL SECTIONS This subsection summarizes the structural design of representative scismic Category I structural elements in the auxiliary building and shield building. These structures are listed below and the corresponding location numbers are shown on Figure 3}i.51. The basis for their selection to this list is also provided for each stmeture.

(1) South wall of auxiliary building (column line 1), elevation 36'-6" to elevation 180'-0". (This exterior wall illustrates typical loads such as soil pressure, surcharge, temperature gradients, seismic, ami tomado.)

(2) Interior wall of auxiliary building (column line 7.3), elevation 66'-6" to elevntion 160'-6" (This is one of the most highly stressed shear walls.)

(3) West wall of main control room in auxiliary building (column line L), elevation 117' 6" to elevation 153'-0". (This illustrates design of a wall for subcompartment pressurization.)

(4) Nonh wall of MSIV east companment (colurnn line i1), elevation i17t6" to 153'-0",

(The main steam line is anchored to this wall segment.)

(5) Shield building cylinder at elevation 180t0".

Revision: 20 (Draft)

[ Westingh00Se 311-8 January 9,1998

www Appendia 3ll A:siliary fluilding Critical Sections (6) Roof slab at elevation 180'-0" adjacent to shield building cylinder, Ghis is the connection between the two buildings at the highest elevation.)

(7) Floor slab on metal decking (elevation 135' 3")

(This is a typical slab on metal decking and structural steel framing.)

(8) 2'-0" slab in auxiliary building (tagging room ceiling at elevation 135'-3")

(This illustrates the design of a typical 2'-0" thick concrete slab.)

(9) Flimed floor in the main control room at elevation 135'-3" (This illustrates the design of the finned floors.)

(10) Shield building roof /PCCS wa'cr storage tank

(~Ihis is a unique area of the roof and water tank.)

(11) Shield building roof to cylinder kication at columns Ohls is the junction between the shield building roof and the cylindrical wall of the shleid building.)

(12) Divider wall between the spent fuel pool and the fuel transfer canal. (Blis wall is subjected to thennal and seismic sloshing loads) 311.5.1 Shear Walls Structural Description Shear walls in the auxiliary building vary in stie, configuration, aspect ratio, and :unount of n'inforcement, The stress levels in shear walls depend on these parameters and the seismic acceleration level. The range of these parameters and the stress levels in various regions of the most severely stressed shear wall are described in the following paragraphs.

The height of the major structural shear walls in the auxilhtry building nmges between 30 to 120 feet. The length ranges between 40 and 260 feet. The aspect ratio of these walls (full height / full length) is generally less than 1.0 and often less than 0.25. Therefort, these walls fall within the cate;ory of low rise shear walls. The walls are typically 2 to 5 feet thick, and are monolithically cast with the concrete floor slabs, which are 9 inches to 2 feet thick.

Exterior shear walls are several stories high and do not have many large openings. Interior shear walls, however, are discontinuous in both vertical and horizontal directions. The in-plane behavior of these shear walls, including the large openings, is adequately represented in the analytical models.

Bie shear walls are used as the primary system for resisting the lateral loads, such as earthqual'es. The auxiliary building shear walls n't also evaluated for flexnte due to the out-of-plane loads.

Revision: 20 (Draft)

T Westinghouse 311-9 January 9,1998

^

,r 7

- APrendiz 3H Atsillary B:llding Critic:t Sections j

-u -

.l

- Design Approach The auxiliary- building shear walls are designed 'to withstand the _ loats specified in subsection 311.3.3, Beside dead, lis, and other nonnal operating condition lotuls, the

- following loads am considered in the shear wall design:

Seismic loads - a

- 11.c SSE loads for the wall att obtained fmm the seismic analyses of .;

auxiliary /shleid buildings that are described in subsection 3H.4, i

-- Calculations are perfonned by considering shear wall segments bounded by the floors below arxl above the segment and the adjacent walls perperxticular to, on both sides of, the segment under consideration. Appmpriate boundary conditions are assumed for the four edges of the segment. Natural frequency of wall .

segments are detennined using text book fonnulas for the frequency of plate structures. Corresponding spectral acceleration is detennined from the applicable .-

response spectrum.

2

- Exterior walls, below grde level, are also evaluated for dynamic carth pressure exened during an SSE.

Accident pressure load Shear walls of the main steam isolation valves (MSIV) rooms are designed for 5 pounds per square inch (psi) differenilat pressure acting in conjunction with the scismic loads. Member forces due to accident pressure and SSE are combint d by absolute sum.

'lhe main contml room wall of the cast MSIV compartment is evaluated for the perssure and the jet load due to a postulated main steamline bmt.

Tornado load l'or exterior walls above gnule level, tomado loads are considered.

The design temperatures for thennal gradient are included in Table 3H.S-1.

The shear walls are designed for the load combinations, as applicable, con;ained in Table 3.8.4 2. the wall sections are designed in accordance with the requirements of AC1349 90, Revision: 20 (Draft)

W Westinghouse 311-10 January 9,1998

- - . ~ . - - . .

~ Appendix 311 A siliary it:liding Critical Section, 311.5.1.I' Exterior Wall al Column 1.ine 1 The wall at column line 1 is the exterior wall at the south end of the nuclear island.- The reinforced concrete wall extends from the top of the basemat at elevation 66t6" to the roof at elevation 180t0". It is 3'-0" thick below the grade and 2' 3" thick above the gnule.

The wall is designed for the applicable loads including dead load, live load, hydrostatic load, lateral soll pressure loads, seismic loads, and thennal losis. As shown in Figum 311.5 2, the wall is divided in 12 segments for design purpose. Table 311.5-2 provides the listing and magnitude of the various design loads. Table 311.5+3 presents the goveming load combination for each wall segment and the details of the wall reinforcement. The actual reinfortement provided is compared to the required irbar area for each wall segment. Figure 311.5-3 shows the typical minforcement for the wa'.1 at column line 1.

311.5.1.2 WaW at Column I,ine 7.3 The wall at column line 7.3 ls a shear wall that connects the shield building and the nuclear island exterior wall at column line I. It externis from the top of the basemat at elevation 66t 6" to the top of the roof. The wall is 3 feu thick below the grade at elevation 100'-0* and 2 feet thick above the grade. Out of-pla'le lateral support is provided to the wall by the floor slabs on either side of it and the roof at the top.

Wall 7.3 is designed for the applicable loads desenbed in subsection 3H.3.3.

For various segments of this wall, the corresponding goveming load combination and associated design loads are shown in Table 311.5-4.

Table 311.5 5 presents the details of the wall reinforcement. The actual reirifortement pmvided is compared to the required reinforcement area for each wall segment. Typical wall reinforcement is also shown in Figure 311.5-4 311.5.1.3 Wall at Column Line L 1he wall at column line L is a shear wall on the west side of the Main Contml Room. It extends fmm the top of the basemat at elevation 66'-6" to the top of the roof. The wall is 2 feet thick. Out-ef plane lateral support is pmvided to the wall by the floor slabs on either side of it and the roof at the top. The segment of the wall that is a past of the main contml mom boundary is from elevation i17t6" to elevation 135t3".

The auxiliary building design loads are described in subsection 311.3.3, and the wall is designed for the applicable loads, in addition to the dead, live and seismic loads, the wall - ,

~ is designed to withstand a 5 pounds per square inch pressure load due to a pipe break in the MSIV room even though it is a break exclusion area. This wall segment is also designed to withstand a jet load due to the pipe break.

The goveming load combinallon and asso+%I design loads are shown b Table 311.5-6.

Resision: 20 (Draft)

! W Westinghouse 311 11 January 9,1998 -

li!MNillt Appendix 311 Armillary Itilding Critic 1 Sections n-Table 311.5 7 presents the details of the wall re'nforcement. The actual reinforcement provided is wmpamd to the required reinforcement area for each wall segment.

311.5.1.4 Wall at Column Line 11 The north wall of the MSIV cast compartment, at colmnn line 11 between elevation 117'-6" and elevation 153'-0", has been identified as a critical section.

The segment of the wall between elevation 117'-6" and elevation 135'-3"is 4 feet thick, and several pipes such as the main steam line, main feed water line, and the start-up feed water line are anchored to this wall at th, interface with the turbine building.

The wall segment from elevation 135' 3" to elevation 153'-0"does not provide support to any high energy lines, and is 2 feet thick. This portion does not have to withstand reactions from high energy line breaks.

The wall is designed to withstand loads such as the dead load, live load, seismic load and t'te thennal load. The MSIV room is a bmak exclusion area, but the design also considered the loads associated with pipe rupture in the MSIV room, such as compartment pressurization, jet load, and the mactions at the pipe anchors. The loads on the pipe anchor include pipe rupture loads for breaks in the turbine building.

The wall structure is analyzed using three dimensional finite element analyses. Analyses are perfomied for individual loads, and design loads am detennined for applicable load combinations from Table 3.8.4-2. The design is perfomled for the enveloping cares for critical regions.

General features of the wall reinforcement are shown in Figure 311.5-5.

311.5.1.5 Shield lluilding Cylinder at Elevation 180'-0" 1he thickness of the cylindrical portion of the shield building wall is 3 feet.

The wall is designed for the applicab!e loads desenbed in subsection 311.3-3. A detailed finite element arnlysis is perfonned to detennine the design forces. The amount of minforcement in horizontal and vertical directions provided on each face is same. Typical minforcement fmm elevation 200' 0" to 160'-6", on each face, is as follows:

Elevation 200'-0" to 180'-6": Requimd horizontal reinforcement = 3.45 inch2 /ft.

Provided horizontal reinforcement = 3.81 inch'/ft.

Required vertical reinforcement = 3.71 inch'/ft.

Provided vertical reinforcement = 3.81 inch'/ft.

Resision: 20 (Drah)

W Westinghouse 311-12 January 9,1998 l

Appendis 311 A:sillary fluilding Critical Sections Elevation 18016" to 160t6": Required hodzontal reinforcement = 3.68 inch'/ft.

Provided horizontal reinforcement = 3.81 inch'/ft.

Required venical rtinforcement = 4.00 inch'/ft.

Pmvided vertical reinforcement = 4.50 inch'/ft.

The design of the shield building roof is described in 311.5.6.

311.5.2 Composite Structures (Floors and Roof)

The floors consist of a concrete slab on metal deck, which rests on structural steel floor beams. Sevend floors in the auxiliary building are designed as one-way reinforced concrete slabs supported continuously on steel beams. Typically, the beams span between two reinforced concrete walls. The beams are designed as composite with fonned metal deck spanning perpendicular to the members. Unshored construction is used. For the floors, beams are typically spaced at about 6-feet intervals and spans are between 16 feet and 25 feet.

For the roofs, beams are typically spaced at 12 to 14 foot intervals.

Structural Description A typical layout of these floors is shown in Figure 311.5-6. The metal deck rests on the top Hange of the structural steel floor beam, with the longitudinal axes of the metal deck ribs and floor Icams placed perpendicular to each other. The depth of the ribs for 9-inch concrete thor slabs and 15-inch deep concrete roof slabs are 3 inches and 4.5 inches respectively. The concrete slab is tied to the structural steel floor beam by shear connectors, which are welded to the top flange of the floor beam. The concrete slab and the floor beams fonn a composite floor system. For the design loads after hardening of concretc. .he tnnsfonned section is used to check the stresses.

The construction sequence is as follows:

The structural steel floor (floor beam, metal &ck, and shear connectors) is fabricated in the shop, bmught to the floor location, and placed in position. In some cases, the beams rud deck are preassembled and placed as a module.

The metal deck is used as the fonnwork, and concrete is poured on the metal deck.

Until concrete hardens, the load is carried by the m*tal deck and the steel floor beam.

During concreting, no shoring is provided.

Design Approach The floor design considers the dead, live, construction, extreme environmental, and other applicable loads identified in Section 311.3 3. The design floor loading includes the equipment attached to the floor. The end condition for the steel beams is simply supported, Revision: 20 (Draft)

T Westinghouse 311-13 January 9,1998

l Appendix 311 Ausillary llullding Critical Sections or continuous. The seismic load is obtained using the applicable floor acceleration response spectrum (7 percent damping for the SSE loads).

The load combinations applicable to the design of these floors are shown in Tables 3.8.4-1 and 3.8.4 2. The design of the floor system is performed in two parts:

Design of structural steel beams The stmetural steel floor beams are evaluated to withstand the weight of wet concrete during the placement of concrete. The composite section is checked for the design loads during nonnal tuid extreme envirorunent conditions. Shear connectors are also designed.

Design of concrete slab The concrete slab and the steel reinforcement of the composite section are evaluated for nonnal and extreme envimrunental conditions. The slab concrete and the reinforcement is designed to meet the requirements of American Concrete Institute standant ACI 349-90 " Code Requirements for Nuclear Safety-related Structures."

The slab design considers the in-plane mul out-of-plane sels.nic forces. The global in-plane and out-of-plane forces art obtained from the ressmse spectmm analysis of the 3D finite element model of the auxiliary and shield buildings. The out-of plane seismic fortes due to floor self-excitation am detennined by hand calculations using the applicable venical scismic response spectrum and slab frequency.

311.5.2.1 Roof at Elevation 180' 0", Area 5 (Critical Section is between Col. Lines N & K 2 and 3 & 4)

The layout of this segment of the roof is shown in Figure 311.5-7 as Region "11." 1hc concrete slab is 15 inches thick, plus 4.5-inch deep metal deck ribs. It is composite with 5 feet deep plate girders, spaced 14'-2" center to center, by using shear cormectors. The ginler flarges are 20" x 2" and the web is 56" x 7/16." The girders span approximately 64 feet in the north south direction imd are designed as simply supported. The concrete slab between the ginlers behaves as a one-way slab and is designed to span between the giniers.

The ruof ginlers are designed for dead and live loads, including constmetion loads (with wet concrete) with :imple support end conditions. A one-third increase in allowable stress is permitted for the construction load combit'ation.

1he girders am also evaluated as part of the comp > site beam after drying of concrete. The composite roof structure is designed to withstand dead and .ive load / snow load, as well as the wind, tomado and seismic loads.

Revision: 20 (Drail)

M WestinghollSe 31144 January 9,1998 l l

Appendix 311 A sillary llullding Critical Sections A typical c(umection of the niof slab to the shield building is shown in sheet 2 of Figure 311.5-7.

Governing Load Combination (Roof Ginier):

Combination Number 3 - Extreme Environmental Condition Downward Seismic Acceleration Bending htoment = 5402 kips-ft.

Corresponding Stress = 18.3 ksi Allowable Stress = 33.3 ksi Shear Force = 339 kips Corresponding Stress = 12.9 ksi Allowable Stress = 20.1 ksi Governing Load Combination (Concrete Stab):

(Parallel to the Giniers)

Combination Numbers 3 - Extreme Environmental Con 0aion Upwan! Seismic Acceleration Bending Atoment = 2152 kips-ft. (per 14.17 ft width of the slah)

In-plane Tension = 71.9 kips (per foot width of the slab)

In-plane Shear = 66.1 kips (per foot width of the slab)

Reinforcement (Each Face)

Required = 1.17 sq. inch /ft.

Provided= 1.56 sq. inch /ft.

(Perpendicular to the Gin!crs)

Combination Numbers = 3 - Extreme Envirotunental Condition Bending htoment = (-) 14.6 kips-ft. (per foot width of the slah)

Bending htoment = 15.2 kips-ft. (per foot width of the slab)

Revision: 20 (Draft) 3 Westilighouse 3H15 january 9,1998

Appendis 311 Auxiliary llullding Critiell Sections (Due to lhennal Gradient)

In-plane Tension = 71.9 kips (per foot width of the slab)

In planc Shear = 66.1 kips (per fcc.; width of the slab)

Reinfortement (Each Face)

Required = 1.92 sq. inch /ft.

Provided= 3.12 sq. inch /ft.

311.5.2.2 Floor at Elevation 135' 3", Area 1 (lletween Column Lines M and P)

The design of the floor is shown in Figure 311.5-6. The concrete slab is 9 inches thick, plus 3-inch deep metal deck ribs. The floor beams are typically W14x26.

The floor beams are designed for construction load (with wet concrete) with simple support end conditions. The design k> ads include the dead load and a constmetion live load of 100 pounds per square foot (psi) distributed load plus 5000 pounds concentrated load near the point of maximum shear and moment. A one-third increase in allowable ,

stress is pennitted.

The floor beams are also evaluated as part of the composite beam after drying of the concrete. Ilecause of continuity of rebars into the wall and the connection of the bottom flange to the support embedment, the end s'rpport condition is considered as fixed.

Revision: 20 (Draft) 3 Westkigh0tlS8 311-16 January 9,1998

Appendit 311 ACxiliary Building Critical Sections Goveming lead Combination (Steel Beam):

lead Combination Nomud Condition llending Moment = (-) 64.4 kips-ft.

Corresponding Stress = 16.6 ksi Allowable Stress = 23.76 ksi Shear Force = 25.4 kips Corresponding Stmss = 9.8 ksi Allowable Stress = 14.4 ksi Goveming lead Combination (Concrete Slab):

(Parallel to the Beams) lead Combination 3 Extreme Envirotunental Condition Downwan! Seismic Bending Moment = (+) 47.4 kips-ft. (carried by effective width)

In-plane Shear = 18.1 kips (per foot width of the slab)

Reinforcement (F.ach Face)

Required = 1.49 sq. inch /ft.

Pmvided = 1,56 sq. inch /ft.

(Perpendicular to the Beams)

Combination Number Nonnal Condition Bending Moment = (-) 8.28 kips-ft. (per foot width of the slah)

Reinforcement (Each Face)

Required = 0.47 sq. inch /ft.

Pmvided = 0.60 sq. inch /ft.

311.5.3 Reinforced Concrete Slabs Reinforced concrete floors in auxiliary building are 24 inch or 36 inch thick. Ecse floors are constructed with 16" or 28" of reinforced concrete placed on the top of 8 inch thick precast concrete panels. The 8" thick precast concrete panels are installed at the bottom to Revisien: 20 (Draft)

[ WestinghollSe 311-17 January 9,1998

Appendia 31I Attillary II:llding Critic:1 Sections r,erve as the fonnwork and withstand the load of wet concrete slab. 'lhe main reinforcement is provided in the precast panels which are connected to the concrete placed alove it by shear

' reinforcement. The precast panels and the cast-in-place concrete act together as a composite reinforced concrete slab. Examples of such floors are the Tagging Roorn ceiling slab at elevation 135 ft 3 inches in Area 2, and the Area S/6 elevation 100t0" slah between column lines 1 & 2.

311.5.3.1 Tagging Room Ceiling Design dimensions of the Tagging Room Ceiling are as follows:

Room Size: 16'-0" x 1 I t10" Boundary Conditions: Fixed at Walls J and K Clear Span: 16t0" Slab Thickness: Total = 24 inches Precast Panel = 8 inches Cast-in Place = 16 inches lhe two precast concrete panels, each Stl1" wide and spanning over 16t0" clear span, are installed to serve as the fom1 work.

Design of the Precast Concrete Panels:

Goveming Load Combination = Construction Design Bending hfoment (htidspan) = 14.53 ft-kip /ft.

Bottora Reinforcement (E/W Direction) Required = 0.51 in'/ft.

Bottom Reinforcement (E/W Direction) Provided = 0.79 in'/ft.

Top Reinforcement (E/W Direction) Required = (Alinimum required by Code) 2 Top Reinforcement (E/W Direction) Pmvided = 0.20 in /ft.

Top and Bottom Reinforcement (N/S Direcilon)

Required = (htinimum required by Code)

Top and Bottom Reinforcement (N/S Direction)

Pmvided = 0.20 sq. In/ft.

Revision: 20 (Draft)

[ W85tifigh00S6 311-18 January 9,1998

Appendis 3fl Anillary llullding Criticil Sections Design of 24 inch Thick Slab:

Goveming lead Ccunbination = Extreme Envirorunental Condition (SSE)

Design Bending Moment (N/S Direction)

Midspan = 12.31 kips ft/ft Design in-plane Shear = 34.14 kips /ft Design in-plane Tension = 18.99 kips /ft 2

ficttom Reinforcement (E/W Direction) Required = 0.64 in /ft Bottom Reinforcement (E/W Direction) Provided =. 0.79 in'/ft Design Bending Moment (N/S Direction) at Support = 24.63 kir.-ft/ft Design in-plane Shear = 34.14 kips /ft Design in-plane Tension = 18.99 kips /ft 2

Top Reinfortement (E/W Direction) Required = 0.78 in /ft Top Reinforcement (E/W Direction) Provided = 0.79 in'/ft Design Bending Moment (N/S Direction) = 7.28 kips ft/ft Design In-plane Shear = 34.14 kips /ft Design In-plane Tension = 16.99 kips /ft Top and Bottom Reinforcement (N/S Direction) 2 Required = 0.64 in /ft Top and Bottom Reinforcement (N/S Direction)

Piuvided 2

= 0.79 in /ft 311.5.4 Concrete Finned Floors The ceilings of the main control room, and the instrumentation and control nxnns in the auxiliary building are designed as finned-floor modules. A typical floot design is shown in Figure 3.8.4-6. A finned floor consists of a 24-inch-thick concrete slab poured over a stif fened steel plate ceiling. The fins, welded to stiffen the steel plate, are rectangular sections perpendicular to the plate. Shear studs are welded on the other side of the steel plate, and Revision: 20 (Draft)

[ We5fingh00S8 31119 January 9,1993

gifTil i Appendit 311 A:siliary llullding Critical Sections no,v..

the steel and concrete act as a comp > site section. The fins are exposed to the envinnunent of the nxnn mal enhance the heat absodiing capacity of the ceiling. Several shop fabricated steel panels, cut to nunn width arul placed side by $1de perpendicular to the nunn length, are us:xt to construct the stiffened plate ceiling in a rmxtularised fashion The stiff ened plate with fins is designed to withstand construction loads prior to concrete hanlening, The main c<nitrol nxnn ceiling fin thxir is designed for the dead, live, and the seismic loads.

De finned ikuir structure is evaluated for the load c<nnbimitions listed in TrM.'s 3.8.41 mui 1.8.42.

Dedgn Methodology 1he finned ihxirs are designed as reinforced concrete slabs in accontance with ACI Standant 349. For positive bending, the steel plate is in tension. The steel plate with fir, stiffeners serves the function of bottom rebarx. For negr.tive berating, the pitential for buckling due to c4nnpression in thL clement is checked by using the criteria of American National Standants Imtitte/Arnetican Institute of Steel Construction standants ANSI /AISC N690 84. Twisting, mal therclam lateral buckling of the stiffener, is restrained by the concrete.

1hc linned thxirs resist vertical mal in plane fortes for lxith nonnal mal extreme loading conditions. For positive bending, the concrete above the neutral itsis carries ceinpressive stresses arul the silffened steel plate resist: tension. Negativr bending compression is resisted by the stiffened plate mid tension by top rebars in the concrete.1hc neutnd axis for negative bending is hicated in the stiffened plate section, mut the concrete in ic'ision is assumed inactive. Ilorimntal in plane fortes are resisted by the stif feneti plate aint longitudinal arbars.

Mhilmurn top reinforcernent is pnivided in the slah in each direction for shrinkage mal tem "ature crack contrui, In addition, top reinforcement kicated parallel to the stiffeners is usi tension reinforcement in negative bending. The stif fened plate pnivides crack control cap .y for the tuttom tJ the slab in the tuuisverse direction.

Composite section propenics, based on an all steel transfonned section, as detailed in Section Q1.11 of ANSI /AISC N690-84, are used to check the following:

Wehl strength between stiffener mal the steel plate

+

3 pacing of the shear studs for the compostic action 1he stilfened plate alone is designed to trsist all construction loads prior to the concrete haniening. The plate is checked against the critesia (or bending and shear, specified in ANSI /AISC N690 84, Sections Ql.5.1.4 mul Ql.5.1.2. In addition, the wehl between the stiffener atal the steel plate is checked to satisfy the c(xle requirements.

Revision: 20 (Draft) 3 W8dingh00S8 311 20 January 9,1998

Appendis 311 A:siliary llullding Celaical Sections 3115.4.1 MCR Ceiling (lloor at Elevation 135'.2")

The design of the bottuu plate with fins is govenied by the construction load. For the cornp9 site floor, the design forces used for the evaluation of a typical 9 inch wide strip of the slah are as follows:

Matirnurn bending inoinent = 134.5 klps ft

Maxirnuin shear force = 14.9 Lips lhe design evaluation results are summarlied below:

'ihe actual area of the tension steel is 9.0 in', which pnwides a design strength of 518.5 kips ft tending inoinent capacity.

The design shear strength is 23.22 Lips.

1he shear studs are spaced 9 inches c/c,t i' both directions. 'Ihe calcult si required spacing is 15.7 inches.

Re51sion: 20 (l> raft)

W Westinghouse 31121 January 9,1998

{lliifl{ APPendit 311 A:sillary fluilding Crilled Nettions KAW -

311.5.5 Structural Modules Structural imx!ules are used for part of the south side of the auxiliary building. These structural matules are structural elements built up with welded steel structural shapes and plates. The imulules consist of st el faceplates connecttd by steel trusses. The pdmary purpose of the trusses is to stiffen and hold together the faceplates during handling, erection, and nuicrete placement.1he nominal thickness of the steel faceplates is 0.5 inch except in a few hical arras. The n(nuinal spacing of the trusses is 30 inches. Shear studs are welded to the inside faces of the steel faceplates. Face plates are welded to adjacent plates with full penetration welds so that the weld is at least as simng as the plate. The structural wall snodules are anchored to the concrete base by reinforring steel dowels or other types of cosmections embc(kled in the reinforted unicrete below. After erection, concrete is placed between the faceplates.

These modules include the spent fuel pool, fuel trrsfer canal, and cask loading and cask washdown pits. The structural m(slules are similar to the structural modules for the containtnent intemal structures (see subsection 3.8.3). Figure 3.8.4 5 shows the hication of the structural auxlules in the auxiliary building. The structun modules extend from elevation 66' 6" to elevation 135' 3",

The loads and load combinations applicable to the structural m(xlules ili the auxiliary building are the same as for the containment intemal structures (subsection 3.8.3.5.3) except that there are no ADS 1.or pressure loads due to pipe biraks.

1he design ineth(xtology of these modules in the auxiliary building is similar to the design of the structural linxlules 11i the contaltimerit iritemal structures described Iri subsectioin 3.8.3.5.3.

311.5.5.1 West Wall of Spent Fuel Pool Figurr 311.510 shows an elevation of the west wall of the spent fuel pool (colurun line L 2), and element number

In the finite element nuxlel 1hc wall is a 4 feet thick concrete filled structural wal' module.

A finite element analysis of the spent fuel bulkling tuodule is perfonned for 3eismic, thennal and hydrustatic loads with the following assumptions:

1he analysis m(xlel includes the structure between Lines 2 and 4, Lines I and N, arul between El 66' 6" and 135' 3", arul is fixed at the base. There is no suppcrt at elevation 135' 3"

'lhe seismic input consists of floor trspese spectra derived frorn the spectra for the skior at El 135' 3", which are conservatively applied at the basemat level as ground response spectra.

Revision: 20 (Draft) 3 Wftstinghouse 311 22 January 9,1998

Appendis 311 ACsiliary fluilding Critical sections m.,,

1he thennal loads are app!!cd as lineady varying temperatures between the inner tuid outer faces of the widis and thors.

'Ihe hydrostatic loads are applied to the spent fuel pot walls and ikors, which is considered full with water. This provides the loads for the ucsign of the divider wall.

'Ihe seismic sloshing is imxteled in the spent 16 :1 pol.

The concrru fille.t sta etural wall imxtules are designed as reinforced concrete stmetures in accontain " th the reqalrements of ACI 349. The face plates are treated as reinforcing steel.

Methods of analysis are based on accepted principles of structural mechanics and are consistent with the geometry and toundary conditions of the structures. Iloth computer codes and h:uid calculations are used.

Table 311.5 8 shows the magnitude of typical design loads, load combinations, and the required and provided plate thickness for certain critical kications. The steel plates are generally half inch thick.1hc plate thickness is increased close to the bottom of the gate through the t 'l where the opening results in high kical member forces. 'Ihe first part of the table shows the snember forces due to individual loading. The lower part of the table shows the goveming load combinations.1he sicci plate thickness required to resist mechanical leads is shown at the bottom of the table as well as the thickness pn'vhirt. The maximum principal stress for the load combination including thennal is also wh; ated. 'I this value exceeds :he yleid stress at temperature, a supplemental evaluation is perfonnen. For these cases, the maxhnum stress intensity range is shown together with the allowable stress intensity range which is twice the the ylekt stress at the temperature.

311.5.6 Shield hnilding rimf The shield building nof is a reinforced concir:e shell supponing the passive containment cooling system tank tuut air diffuser. Air intakes are hicated at the top of the cylindrical pirtion of the shield building. The conical mof supports the passive contaitunent cooling system tank as shown in l'igure 3 H.4 7. The design of critical areas is discussed below.

'Ihese areas include the tension ring at the connection of the conical nof to the cylindt' cal wall, the columns between the air inlets just below the air inlets, and the exterior wall of the passive containment cooling system tank.

311.5.6.1 Tension ring

'the connection between the conical nof and the shiehl building glindrical wall is designated as the tension ring, it spans as a bearn across the air inlets. The goveming load for the tension ring is axial tension. The maximum tension is atout il(X) kips under nonnal operating loads. SSil seismic hiads result in maximum axial loads of about 18(X) kips. The combined load ranges fium 2900 kips tension to 8(X) kipc compression. The maximum axial tension irsults in a reinforcement stirss of 34 xsi. The reinforcement will also see tensile stresses due to other member forte comIonents, primarily torsion and bending atout the Revision: 20 (I) raft)

T Westinghouse 311 23 January 9,1998

l%l Appendis 311 A:sillary llullding Critic:1 Sections m-horinnital axis. 'lhe inaxirnuni axial cornpression results in a concrete cornpressive stress of 3F,0 psi. This is less than 10 percent of the concrete cornpressive strength. 'Ihe ring is designed as a tenslori inernber; shear stirrups are pruvided to carry the shear and torsion without taking credit for concrete shear strength. The reinforcenient is shown in Firure 311.5-9. 'Ihe reinforectnent required and provided is surninarized in sheet I of Table 311.5 9.

311.5.6.2 Colurnn (shear wall) between air inlets

'lhe colutun between the air inlets has plan dirnensions of 36" x 138" arxl is 60" high. Its priinary loading is vertical load due to dead arnt scisinic loruls arul horitoritid scistnic shear, it is designed as a low rise shear wall. The axial cannpression is about 1200 kips urxler nonnal operating loads. SSh selsinic loads result in inaxirnuin axial loruls of about 17(X) kips.

The cornbined load ranges Inun 2900 kips cornpression to 500 kips tension. The inaxlinuin horirontal shear is 2200 kips in plane ruxi 800 kips out-of plane (D.L. = 300, SSF = 500).

'Ihe 2900 kips c4nnpression contsponds to an axial cornpressive stress of about 6 XI psi.

1hese loads arnt the associated bernling nionients result in a inaxlinurn concrete conipressive stress of 1400 psi asul a inaxirnuin concrete tensile stress of 800 psi at the base of the colurnri assurning gross concrete section pnipct,les. 'the reinforectnent is shown in the Figure 311.5 9.

'!he reinforcernent required and pnivided is surninarized in sheet 2 of Table 311.5-9.

311.5.6.3 lhterior wall of the paulte containtnent cooling system tank

'!he extedor wall of the passive containtnent cooling systern tank is two feet thick. 'Ihere is a stainless steel liner on the inside surface of the tank. 'Ihe wall liner consists of a plate with stif feners and welded studs on the concrete side of the plare. Leak chase channels are provided over the liner welds. 'lhe reinforecruent in the concrete wall := designed without taking credit for the strength provided by the liner. 'Ihe goveming loads for design of the exterior wall are the hydn> static pressure of the water, the in plane anni out-of plane seismic response, and the ternperature gnatient across the wall. The reinforcement required ruid pnivided is sununarized in sheet 3 of Table 311.5-9.

l(evision: 20 (Draft)

[ Westinghouse 31124 January 9,1998

Appendis 3ll Asillary llullding Critical Sections Table 311.5-1 NUCl EAR ISLAND: DESIGN TEMI'ERATURES FOR TilERMAL GRADIENT Structure i nad Temperature (*1') Remark (Outside) (Inside)

I'CS Tank Walls Nonnal 1hennat, T. -40 +40 -

+115 +40 (Outside) (Inside)

Roofs and Ihterior Nonnal 1hennal, T. +10 +70 -

Walls Alwive Gnide +100 +70 Accident 1hennal T4 10 +132 MSIV nxun exterior walls and roof (Side 1) (Side 2) interior Walls / Slabs Nonnal lhennal, T. N/R !J/R -

Accident Thennat, T4 +70 +132 MSIV nxnn interior walls and slabs lhterior Walls Below Grade Nonnal lhermal. T, N/R N/R -

Accident 1hennal, T4 N/R N/R -

llawmat Nonnal lhennal. T,, N/R N/R -

Accident 1 henna!, T4 N/R N/R -

Shield flullding (Outshle) (Inside) -

(Iletween Upper Annulus Nonnal Thennal T. 10 +70 anti Auxiliary llullding) +100 +70 Accident Thermal, T. 10 +132 MSiv nxun wall N/R N/R Rest of wall huttil 1, NM means touts due to a thennal tradient are not requirnt to be considensl.

2. Itased on ACl 349#0 (Appenths A), the base temperatuie for the mnstruction is assumed to be 70'F.

Revision: 20 (Draft) 3 Westinghouse 311 25 January 9,1998

lilll"'llI3 Appendis 311 ACsillary Hullding Critical Sections Table 311.5 2 (Sheet 1 of 2)

Esterior Wall on Column Line 1 FORCES AND MOMENTS IN CRITICAL IA) CATIONS (See Figure 311.5 2 l'or Locations) newa.. e ...t n, s a..e . . .+ >

Lesd t and begegenst Wat( $$CTION WeebaenedEL4J W 'd i W filfDf^ - ~

w.. 1 i . . . e t . . ,

amow o .et .or. s ie .. 4. -.= s ie e ce 4u 4n ou e s.

%..,~, . . o0 t es .e , n .e 4. i se 9* en -te aa m re LL wr e6 c. 4 cm ose e as a se 4 33 4 el 0 08 0 06 4 ft 4 78 t .e ca. ce, en .. 4= 4. i .s t i i. 4a 4a e ,s e ti 1,.

Secuns Water 83 70 tf to FA e0 fl s0 e to .t so it *0 to N e 00 a to mas at to et to .pe en ps an e to .I an to te 80 po 0 00 t to LAf utAL tatt H l'fgfaQ(g a< N.i P. 41 to t e to .ta w .ta no e so 1 to e on 4 to 6 no te idahtlG osaw ne.w ow ser st or u st te os taco $ no eso a re a re Paessee hoe Press 18f 30 N eo 14t t0 +14120 F310 t '0 76 14 65 08 4731 19 79 8 De. tas eme t si so to se tu to .tw to sea on 30 as u ao 43 f t es se -et to tvt Desme+ps 4 to 0 43 1 es 4 Se ier 4 tr 9 48 4 67 OM 4 as wes 6+v t to 1 to -t to -t to e 00 4 to 1 is t 11 1 *e t $$

21196E

f. we.awy 4 as t so it to 11 to .it so s7 40 t is e is 4 to .a to Moment w6 se anchestes tenasun on Die outsue fece of well.

. Ernent we e.gn tw6: sten termen on the sarte f ace of was.

Revision: 20 (Draft)

W Westinghouse 311 26 January 9,1998

Appendit 3ll AChiliar; llallditig Critical .Nections Table 311.5 2 (Sheet 2 of 2)

Exterior Wall on Colurnn I.Ine I FORCES AND MOMENTS IN CRITICAL,l OCATIONS the c6 ewe u,.., p a no 04,s.mw kw kwiii ues Lees wee watn ttrerw - . .-, wait etetrw r- - -

-+

fp t e e to et , 13 9 e it i t _ _.

5 At'LtDQ D wae w re .t se set ter o ss 4W tt.t c es 6M 4 48 4 48 ente hee. esp 4 en 9 6 Po e ri o e w et ._ e ut 0 0

.M LDAD rener LL entsere Pav%ed 1 94 ? to 6 89 to 03 28 31 34 4 23 +4 31 t 48 +143 4 en tese Pavises d ie 4 18 1 44 6 61 FM +# 4 4 29 4 29 4 69 4 63 Hyeoeien, Owswis weie 1 80 0 8 46 t oo 0 0 DM t os 4 6 F *se e av e 9 s0 0 40 6 e c et e s.J 0 0 Ef A AL E3L 04 "1\Lk?d5 p*, e no a to e o no e in e o e ni om o o tit 13WG

- w twe= 37st ta so is so se a sus as tea at an 4is is to is 7e enwve ad Pesea 4 70 0 8 29 f fe 0 0 GH 4H 4M 4M

t. tw tes twe -30 90 9 00 let 4M i 38 0 0 92 0 32 4 18 4 tl byn Sasewpe 4 tt 0 0 0 0 0 e tt # #$ 0 0

.9 e ro 4: e uo u .. no e i ., .,n ,s is .=

n1 tut

t. o, ew4 v ao -u n <n <n .te se e 4n en l et t ot

. un.nt ws .,n ria.i.. i.n on on me wa..a i.e. u .a

. MMW e4 sign intStates tensbn Ort he kwD late N wnN Itevision: 20 (Draft)

T Westinghouse 311 27 January 9,1998

pli""il

'1 Appendis 3ll ACsiliary llullding Critical Sections n,,,,,,

Table '111,5 3 EXTERIOR WAl.1, ON COI.UMN LINE 1 DETAll S OF WAl.I, REINFOR(".a1ENT (Inch'/ fool)

(See Figure 311.5-2 for Locations for Wall Sections)

Required I'tm1ded lated Otwnbmatum 1 Aratio n 1151a1 UunLf.ntal Shtag VrrtKal Renttttal ,% gat Wall Airl10N 1,2,3 0.57 0.80 1.0t h l .01 A l .Oll+ 1.Ol!,41.07 Outside l'ue 4 16 1.07 4.16 1.27 Inside fue 2. l l 1.07 2,67 1.27 Wall hirl10N 4,5,6 0.29 0.40 1.0t h I .OlA 1.Oll* 1.01.,^ 1.0T, Outside i ne 3.05 1.l $ 3.12 1.27 Inside f nee 2.65 1.15 2.67 1.27 Wall AirllON 7,8,9 nt nme 1.0lb I .Ola 1.011+ 1 ')l?,+ 1.0'1, ouimule I we 2.54 2.05 3.12 2.06 Inmede iare 2.16 1.61 3.12 1.69 Wall StrilON 10,11,12 nt none 1.01 o 1 Ola 1.011 + 1.0l?,41.0T. OutsiJe I:me 3.74 2 75 3.74 3.12 Insidi l'ut 2.73 2.31 3.12 2.34

\

nr. Na Required t

Revision: 20 (Draft)

T Westinghouse 311 28 January 9,1998 n

Appendis 311 A:sillary Itullding Critical Sections Table 311.5-4 INTERIOR WAI.1, AT COI UhlN I.INE 7J FORCES AND AIOhlENTS IN CRITICAL,l OCATIONS (Units: Kips, Pt.)

I,nad Combination hl hly his, T Ty Ts, l' rom Hmd to Eleistion 135' 3" D + L + E, + T4 78.6 90.7 21.2 78.6 118.9 130.8 Elesation 135' 3" to 117' 6" U.9 D + E, 18.2 10.4 7.7 37.4 91.9 1N.6 Elesation 117' 6" to 100' 0" (Iletween Shield llidg. and 1.ine J) 0.9 D + I!, 7.4 14.5 3.7 29.3 84.7 122.4 Ele $ation 117' 6" to 100' 0"(Iletween I,ine J and I.ine 1) 0.9 D + E, 19.2 11.1 3.2 13.1 18.0 57.6 Eleintion 100' 0" to N2' 6" U.9 D + E, 25.0 12,0 6.3 27.6 80.3 116.6 Eleiation H2' 6" to 66' 6" 0.9 D + E, 9.8 13.6 4.4 20.7 39.3 100.2 Note: X is along horisontal direction and Y is in the vertical direction Resision: 20 (Draft) 3 Westinghouse 311-29 January 9,1998

Appendis 3ll Artillary llulld:ng Critical Sections TaNe 311.5 5 INTERIOR WAI,L ON COLUMN LINE 7.3 DETAILS OF WAl L REINFORCEMENT Wall Segment Location Reinforcement on Each I' ace, s.q.in/ft.

Required I'rmlded I rom Roof to EL 135' 3" llorizontal 2.50 2.56 Vertical 3.03 3.12 Elevallon 135'.3" to 117' florizontal 1.44 1,56 Vertical 1.85 2.00 Elevation !!7'-6" to 100' llorizontal 1.32 1.56 (Iletween Sh.Illdg. and J Vertical 1.92 2.00 Elevation 117' 6" to l(K)' ilorizontal 0.73 1.(K)

(!!ctween Line J and Line I) Vertical 0.69 1.(K)

Elevation 100'41" to 82*-6" lforirontal 1.26 1.56 Vertical 1.65 2.00 Elevation H2*-6" to 66'-6" llorizontal 0.87 1.00 Vertical 1.07 1.56 Revision: 20 (Draft)

[ W85tllighouse 311-30 January 9,1998

- =

l Appendis 311 Aalliary flullding Critical Sections Table 311.5 6 INTERIOR WAl.1, AT Col UMN 1 INE 1.

FORCES AND MOMENTS IN CRITICAL LOCATIONS (Units: Kips, Ft.) ,

l l

I,oad combination his hfy hiny T Ty Try IClesallon !!7%" to 135'.3" D + % + P4 + Y, + R, 231.7 299.4 20.1 28.5 66.3 54.0 NO11:: X is Att>N(1 llokt/4WTAl, DIRI LTION AND Y is IN Titi: vt:kTICAl, DIRiri1ON t

Revision: 20 (Draft)

[ W65tilEh00Se 311 31 January 9,1998

Appendis 311 A sillary fluilding Critic:1 Sections Table 311.5 7 Interior Wall on Column I.ine I, Details of Wall Reinforcement Wall Segment I.ocation Reinforcement on Each Face, sq.in/ft.

Required I'rosided Elevation 117*-6" to 135*.3" llorizontal 3.54 3.72 Vertical 4.74 5.12 Shear Reinforcement:

Wall Segment 1.ocation Reinforcement, sq.In/ft.

Required l'rovided Elevation 117*.6" to 135* 3" In cast west direction 0.88 1.2 1

Revision: 20 (Draft) 3 Westingh00$8 311-32 January 9,1998

Appendix 3H Auxillery Building Critical Sections t

i 1

Table 311.5 8 ( t of 5 ) j l

SPENT FUEL POOL WALL TYPICAL DESIGN LOADS AND LOAD COMBINATIONS ELEMENT NO.1218 42 1 l

l S.(T.) SMT,) SMT.,) N %) N M) N. N, l.nad/ C,omb 6w+t he ae 6em 6.e m he .mm Comments

' Dead (D) 0 17 *11 19 1 52 -03%. 3 28 ,

Irva (L)

Hydro (F) 5 02 516 2 23 19 94 148 92 1 47 31 76 7p1 Sa llUll.DINil HOOl' Hl:lNi~OHCEhti:NT SUhthiARY (Tension Ring) hiember force Heinforcement Heinforcernent Helnforcerneat Itatio required presided prosided required /

sq.In'in. lengih sq.in/in. length prmided Attal + bending 38 # 14 bars 0.7 74" Torsion 0.05$ #9 hoop @ 0.45' O.15 0.37 Torsion 4 vertical shear 2 x 0.05$ + 0.19 = 2 legs # 9 hoop 0.42 0.71 0.30 @ 0.4 5*

~~2 # H ties @ 0.9' Torsion + horitontal : hear 2 x 0.055 + 0.081 = 2 legs # 9 hoop 0.33 0.58 0.19 stirrup @ 0.45' 3 # $ des @ 1.H' Notts~~

1) His ratio is calculated imm the interaction diagram far axial had and inornents for the section and does not include the clicct of torsion loading. It is the ratio of tne huds on the interaction surfa::e divided by the design hads for the same ratio of axial loads and inoments.
Revision: 20 (Draft)
W Westinghouse 311-33 January 9,1998
l Appendis 3ll A siliary llullding Critical Sections rable 311.59F(sheet 2 of 3)
Silitt.D llUll, DING HOOF HEINI'OHCEntENT SUhlhiAltY (Air inlet Colurnn) hiember force Heinforcement Heinforcement Heinforcement Natio required provided prmided required /
sq.in/in. height sq.In/in. height prmided Aslal + bending 48 # 11 bars 0.3 "*
Totsion 0.015 #$ hoop at 6" 0.0? 0.30 Torsion + in plane shear 2 x 0.015 + 0.20 = 3 # 7 ties @ 6" U.39 0.77 0.23 Tonion + out-ct platie 0.37 # $ hoop @ 6" 0.56 0.66 khear 9 # $ ties @ 6" Notes
1) This ratio is caiutated from the interaction diagram for axial load and moments for the section and does not include the effect of torsion loading. It is the ratio of the loads on the 'nteraction surface divided by the design kuuls for the stone ratio of antal loads and moments.
~~2) "Ihe vertical reinforcement in the colurnn is provided to meet minimum vertical reirdorcernent requirements for shear walls.~~
Revision: 20 (Draft)
T Westinghouse 311 34 January 9,1998
4 4 ..-,-h. . g - w..-, L - e +- v - , * -- - - .a 4a u s. u.-+-J 4-- d -46MJ. -...L A- h -.
t
)
Appendit 301 Artillar; Building Critical Secekes
+
l l
Table 311.54%(6heet 3 of 3) !
SHIEl,1) Hull,1)ING NOOF MEINFORCEMENT
~~SUMMARY~~
(Esterkw Wall of the l'awhe Containment Cooling Sptein Tank) ;
Wall Segment lacation Releforcernent on each face, aq.in/h i Hequired l'rovided Elevation 275' 4' to 296' llorizontal 1.80 #9 @ 6" 2.00 Vertical 1.78 #11 @ 0A5' 2.55 Elevation 296' to 300' 8' - llortiontal 1.05 #7 0 6" 1.20 Vertical 0.R$ #11 @ 0.45' l.27 1
b
-k..-,
Revision: 20 (Draft)
W Westinghouse 311 35 January 9,1998
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F 0- 0 0 0 0 0 0 0 0 0 Figure 311.51 (Sheet 1-of 3)
Nuclear Island Critical Sections Plan at El.135'.3"
,. Revision: 20 DraN .
4? January 9,1998 ,
~~3. Dei.lgn a,f Structures Components, a r ulpment, and Spiem.~~
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Nuclear hiand Critical Sections I'lan at El 180'.0" Rulsion: 20 January 31,1998 [ b 11ngh00$8 311 17
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Nuclear Island Critical Sect. ions Section A.A Revision: 20 Draft January 9,1998
"..Y.
WestinghouM 311-F3
I!IMI'i
~~3. Ibign of Structures, Components, I'quipment, and Splems k,~~
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Revision: 20 Draft
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Shield Building Roof Reinforcement at Roof to Wall Connection
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Shield Building Roof Reinforcement at Roof to Wall Connection Revision: 20 Draf1 January 9,1998
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Shield Building Roof Reinforcement at Roof to Wnli Connection
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Shield Building Roof Reinforcement at Roof to Wall Connection Rethion: 20 Draft January 9,1998 W WB5tillgh0USB
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3H-!6F f f

NSD-NRC-98-5515, Provides W Response to FSER Open Items on AP600.Summary of Encl Provided in Table 1.W Status Will Be Changed to Confirm W

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Dear Mr. Quay:

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NSD-NRC-98-5515, Provides W Response to FSER Open Items on AP600.Summary of Encl Provided in Table 1.W Status Will Be Changed to Confirm W

Text

Dear Mr. Quay:

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SUMMARY

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