NSD-NRC-98-5537, Provides W Responses to FSER Open Items on AP600.Summary of Enclosed Responses Is Provided in Table 1.FSER Open Item Number,Associated Oits Number & Status to Be Designated in W Status Column of Oits Is Also Included in Table 1

Provides W Responses to FSER Open Items on AP600.Summary of Enclosed Responses Is Provided in Table 1.FSER Open Item Number,Associated Oits Number & Status to Be Designated in W Status Column of Oits Is Also Included in Table 1

ML20199A714
Person / Time
Site:	05200003
Issue date:	01/22/1998
From:	Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NSD-NRC-98-5537, NUDOCS 9801280063
Download: ML20199A714 (21)

Text

6 Westinghouse - Energy Systems m 355 Electric Corporation Pinsburgti Pennsylvania 15730 0305 DCP/NRCl226 NSD NRC 98 5537 Docket No.: 52 003 January 22,1998-Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 A1TENTION: T. R. QUAY SulljECT: AP600 RiiSPONSE TO FSER OPEN ITEMS l

Dear Mr. Quay:

Enclosure 1 of this letter provides the Westinghouse responses to FSER open items on the AP600. A summary of the enclosed responses is provided in Table 1. Included in the table is the FSER open item number, the associated OITS number, and the status to be designated in the Westinghouse status column of OITS.

The NRC should review the enclosures and inform Westinghouse of the status to be designated in the "NRC Status" column of OITS.

Please contact me on (412 374-4334 if you have any questions concerning this transmittal.

,p u ,

liria A. M it re, Mana(er ef/ A j

Ad\anced 1 t Safety and 1.icensing

,iml 7

Enclosure ec: W. C. Ilufhnan, NRC (Enclosure)

T. J. Kenyon, NRC (Enclosure)

J. M. Sebrosky, NRC (Enclosure)

D. C. Scaletti, NRC (Enclosure)

N. J. Liparulo, Westinghouse 6./o2 Enclosure)

~

k 3 E ,,PDR, h

DCP/NRCl226 NSD NRQ 98 5537 2 January 22,1998 Table I List of FSER Open items included in Letter DCP/NRCl226 FSER Open item OITS Number Westinghouse status in OITS 220.129F (RI) 6434 Confirm W 230.145F 6504 Confirm W 250.35F 6521 Action N l 250.37F 6523 Action N 410.326F 6101 Confirm W l

- Enclosure to Westinghouse Letter DCP/NRCl226 January 22,1998 t

FSER Open item f H Open item 220.129F (OITS #6434) Response Revision 1 Westinghouse should demonstrate the adequacy of using a 6-foot thick foundation mat, especially the foundation mat undemeath the containment vessel. In its position letter dated November 4,1994, the staff offered two options for Westinghouse to consider in resolving this issue: (1) demonstrating that the final foundation mat design can accommodate the effects of soil stiffness variations of hant and soft spots undemeath the foundation mat, and (2) using different foundation mat thicknesses for a foundation mat with uniform soil foundation stiffness (such as rock sites) and for a foundation mat with non unifonn soil foundatian stiffness (such as soil sites with hard and soft spots) and submitting the completed design of each foundation mat thickness for the staff review and approval.

During the meeting conducted on August 4 through 15,1997, the staff reviewed the final design of the 6-foot thick foundation mat. As a result, based on the commitment made by Westinghouse in SAR Section 3.8.4 (Revision 15) that (1) the design and analysis procedures for SC-I stmetures are in acconlance with ACl-349 Code for reinforted concrete structures, and (2) the ductility criteria of ACI 318 Code, Chapters 12 and 21, are considered in detailing, placing, anchoring and splicing of the reinforcing s: eel, three design concems were identified by the staff:

1. According to Chapter 21 (Section 21.3.3.4) of ACI 318-95 Code, stirmps used as shear reinforcement have to be provided with a 135 degree hook at both the top and bottom faces of the foundation mat. However, only stirrups, with 90 degree hook at the bottom face and 135 degree hook at the top face of the mat, were providcd by Westinghouse for resisting shear force. The flexural steel is spaced at 6 inch on centers top and bottom. Etrefore, the provision of 135 degree hooks is not practical ne 6 ft thick foundation mat does not appear to be constructible with such heavy reinforcements.

2. Acconting to the ratio of span to depth, the NI foundation mat should be classified as deep flexural members and be designed for the requirements for deep flexural members (Section Il.8.1 of ACI 349 90). However, based on some test results, the ACI 318 Code Committee determined that errors were identified in the 1983 code (same errors were in the nCl 349 code, because the 349 code is based on the 3 th code) and these crmrs could result in an unconservative design for deep flexural members. As a result, ACI 318-95 was revised to correct these ermrs. For the case of AP600 foundation mat with exterior and interior stiffening walls, the foundation mat should be classified as a continuous deep flexural members (Sections 11.8.1 and 11.8.3 of ACI 318-95).

ACI-318-95 Code (Section i1.8.5) requires that the critical section for shear is to be located at 0.15 times the span length from the support edge with reirdowing steel over the full span and the design should be based on Se: tion 11.8.3. However, Westinghouse did not treat the foundation mat as a deep flexural member. He shear reinforcement used in the design was based on a much reduced shear force at a section which is funher away from the edge at a distance of the effective depth of the mar. De correct amount of shear reinforcement would require the use of larger reinforcing bars which would be spaced at a distance not more than "d/2" throughout the length of the member.

3. De foundation mat calculation was performed using soil stiffness variation in altemate spans.

While this design approach will maximize bending moments in ihe mia span, it will not indicate increases in shear force due 10 soil variation. If the soit variation is such that the soil stiffness is constant over two adjacent spans, and spans on either side are with lower or higher stiffness, the NM" 220.129(RI)-1

O' FSER Open Hem maximum shear force will occur at the wall between the two spans with the greatest stiffness.

This geometry was not considered in the Westinghouse design.

' On the basis of the discussion above, the staff concluded that Westinghouse failed to demonst ste that the proposed foundation mat design is adequate with respect to the previously issued staff position.

'The final design of the foundation mat did not meet certain code regulrements committed to in the SAR.

For the concem of seismic books used for the shear reinforcement (item I above), Westinghouse proposed the use of headed anchors (instead of ;35 degree bends) at both ends of the shear reinforcement (stirrups) during the meeting on = August 4 through 8,1997, Westinghouse also provided, in the meeting, test results published by 'he manufacturer for the staff review. As a result, the staff found that the shear reinforcement with headed anchors is equivalent to the use cf 135 degree bends at both ends of the shear reinforcement"and concluded that this issue is technically resc' <ed.

However, Westinghouse should document this commitment in a Sture revision of the SAR.

With regard to the concems of the use of design code and soil stiffness variation (Items 2 and 3 above), Westinghouse's response and the staff's evaluation are summarized below:

1. . In the submittal dated November 24,1997, Westinghouse provided a draft of S AR Section 3.8.5.5 to commit that the foundation mat below the auxiliary building is designed for -

shear in accordance with the requirements for continuous deep flexural members in ACI 318-95 (Section 11.8.3) Specifically, Westinghouse committed:

(1) The design for shear is based on Sections 11.1 through 11.5 of ACI 349-90 except that the critical section measured from the face of the suppon is taken at a distande of 0.I51,.

(2) Shear strength, V,, is not taken greater than 8(ff)'"b.d when 1/d is less than 2. When #

ljd is between 2 and 5, V,=2/3(IO+1/d)(f/)'"b d (3) Area of vertical shear reinforcement. A , is not less than 0.0015b,s and the spacing of shear reinforcement, s, does not exceed d/2 nor 24 inches.

'4) Shear reinforcement required at the critical section is used throughout the span.

Westinghouse's commitments in the draft SAR meet the requirement of ACI 318-95 and, therefore, are acceptable.

3. In Revision 17 of SAR Section 3.8.5.4.4, Westinghouse stated that the design moments and shears are increased by 20 percent above the required for uniform sites to accommodate the nonuniform sites defined in SAR Section 2.5.4.5. According to the common engineering practice and the staff's review experience, to increase the design moments and shears by 20 percent will accommodate the effects due to nonuniform sites. This issue is considered NO" 220.129(RI)-2

FSER Open Itom 3

m resolved.

On the bases discussed above, the stalf concluded that the concem reganling the design of shear reinforcement for the foundation mat is technically resolved. However, Open item 3.8.5 9 will not be closed unti.' Westinghouse fonnally revises the SAR to document its commitments conceming the use of headed anchors for shear reinforcement and the commitments contained in the submittal dated November 24,1997.

Response: Revision 1 He use of shear reinfwcement with headed anchors is documented in the SSAR revision shown below. De change in design criteria for the design of the bcsemat as a deep slab was included in SSAR subsection 3.8.5.5 in Revision 18.

SSAR Revision:

Revise second paragraph of subsection 3.8.4.6.1.2 asfollows:

In areas where reinforcing steel splices are necessary r .1 lap splices are not practical, I mechanic.I connections (e.g. threaded splices, swaged sleeves or cadwelds) are used.

I I

Headed reinforcement meeting the requirement. of ASTM A970 (Reference 49)is used where

! mechanical anchorage is required. such as for shear reinforcement in the nuclear island basemat I and in the exterior walls below grade.

I i Revise 3.8.5.4.4 as shown below. This also includes revisions discussed during the meetings on 1 January 20 and 2I I998.

I 3.8.5.4.4 Design Summary Report t

l A design summary report is prepared for the basemat documenting that the structures meet the I acceptance criteria specified in subsection 3.8.15.

I I

Deviations from the design due to as procured or as-built conditions are acceptable based on an evaluation consistent with the methods and procedures of Section 3.7 and 3.8 provided the following acceptance criteria are met.

1

• the structural design meets the acceptance criteria specified in Section 3.8 1

I

• the seismic floor response spectra meet the acceptance criteria specified in subsection i 3.7.5.4 I

i Depending on the extent of the deviations, the evaluation may range from documentation of an I

engineering judgement to performar.cc of a revised analysis and design. The results of the i

evaluation will be documented in an as-built summary report by the Combined License I applicant.

e@Me 220.129(R1) 3 l

FSER Opsn itsm .

h

! N.S.445 Design Summary of Critical Sections ne basemat design meets the acceptance criteria specilied in subsection 3.8.4.5. Two critical portions of the basemat are dentined below together with a summary of their design. The boundaries are defined by the walls and column lines which are shown in Figure 3.7.212 (sheet 1 of 12). Table 3.8.5 3 shows the reinforcement required and 'he reinforcement provided for the critical sections.

Basemat between the shield buildine and exterior wall (line 11) and column lines K and L.

Bis portion of the basemat is designed as a one way slab spanning a distance of 23' 6" g

between the walls on column lines K and L The slab is continuous with the adjacent slabs to the east and west. The critical loading is the bearing pressure on the underside of the slab due d to dead and seismic loads. This establishes the demand for the top flexural reinforcement at mid span and for the bottom flexural and shear reinforcement at the walls. The basemat is designed for the bearing pressures and membrane forces from the analyses on unifonn soil springs described in subsection 3.8.5.4.1. The design moments and shears are increased by 20 percent to accommodate the nonuniform sites defined in subsection 2.5.4.5. Negative moments are redistributed as permitted by ACI 349.

De top and bottom reinforcement in the east west direction of span are equal. De I

reinforcement provided is shown in sheets I,2 and 5 of Figure 3.8.5-3. Typical reinforcement I

details showing use of headed reinforcement for sheat reinforcement are shown in Figure 3H.5-1 3.

R;tsemat between column lines I and 2 and column lines K-2 and N This portion of the basemat is designed as a one way slab spanning a distance of 22' 0" between the walls on column lines I and 2. The stab is continuous with the adjacent slabs to the nonh and with the exterior wall to the south. De critical loading is the bearing pressure on

' the underside of the slab de: to dead and seismic loads. This establishes the demand for the top flexura) reinforcement at mid span and for the bottom flexural and shear reinforcement at wall 2. He basemat is designed for the bearing pressures and membrane forces from the analyses on uniform soil springs described in subsection 3.8.5.4.1, The design moments and shears are increased by 20 percent to accommodate the nonuniform sites defined in subsection

! 2.5.4.5. De reinforcement provided is shown in sheets 1,2 and 5 of Figure 3.8.5-3. Typical I

reinforcement details showing use of headed reinforcement for shear reinforcement are shown in i Figure 3H.5 3.

Deviations from the design due to as-procured or as-built conditions are acceptable based on an evaluation consisterit with the methods and procedures of Section 3.7 and 3.8 provided the following acceptance criteria are met.

The structural design meets the acceptance criteria specified in Section 3.8 Tne amplitude of the seismic 000r response spectra do not exceed the design basis lloor response spectra by more than 10 percent

[ Westinghouse 220.129(RI)-4

FSER Open item , . . . .

1 Depending on the extent of the deviations, the evaluation may range from documentation of an 3 engineering judgement to perfonnance of a revised analysis and design.

Add reference 49 I 49. ASTM A 970, " Specification for Welded Headed Bars for Concrete Reinforcement."

l Revise Figure 3.8.5 3 (sheets 4 and 3 of 5) to show additional shear reinforcement.

Add Figure 3H.5-3 also included in response to FSER Open item 220.128F.

l l

l 3 Westinghouse 220.129(R1)-5

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==

3H F6

FSER Open item 230.145F (OITS #6504)

= De design changes made to the CVS as provided in Westinghouse letter NSD-NRC-97 5452 dated November 21,1997, are acceptable to the staff. However, the request for exemption to the definition ,/

of reactor coolant pressure boundary is too broad and, therefore, not acceptable.- The CVS piping within containment is part of the reactor coolant pressure boundary and will not be considered otherwise. An exemption to 10 CFR 50.55a(c) may not be necessary if Westinghouse justifies its proposed altemative classification of the CVS reactor coolant pressure boundary piping under 10 CFR 50.55a(a)(3). Westinghouse should discuss how its classification of the CVS would provide an acceptable level of quality and safety considering the quality, inspection, and integrity criteria for reactor coolant pressure boundary cited in GDCs 14,30, and 32 Westinghouse should also discuss the conformance and exceptions of its classification with regulatory gt:ide 1.26. In addition, Westinghouse should address how the AP600 will be shut down and cooled down in an orderly manner assuming makeup is provided by the reactor coolant makeup syvem only, as discussed in 10 CFR 50.55a(c)(2)(ii). The staff notes that the AP600 would not be able to rely on makeup from the CVS for shut down in an orderly manner since the CVS piping is non-safety related and non-seismic and may not be available due to isolation or oreakage. Westinghouse should include this discussion in the SSAR.

Response

l l Westinghouse is withdrawing the request for an exemption to the definition of reactor coo' ant pressure l boundary. The u:e of nonsafety related piping and components for a portion of the chemical and

( volume control system inside containment that is defined as reactor coolant pressure boundary will be

} justified using the attemate classification requirements of 10CFR50.55a(a)(3). The portion of the l chemical volume and control system that is nonsafety-related is located between the inside contonment i

isolation valves and the valves that provide isolation from the reactor coolant system.

10 CFR 50.55a(3) states the following:

Proposed attematives to the requirements of paragraphs (c), (d), (c), (f), (g), and (h) of this section or portions thereof may be used when authorized by the Director of the Office of Nuclear Reactor Regulation, he applicant shall demonstrate that:

(i) The proposed attematives would provide an acceptable level of quality and safety, or (ii) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The AP600 chemical and volume control system (CVS) is a reliable system, osing high quality industry standards. ANSI B31.1 and ASME Code, Section VIh are used for the construction of the piping, valves, and c6mponents. He nonsafety-related portion of the CVS inside containment will be analyzed seismically. He methods and criteria used for the seismic analysis an: defined in the response to FSER Open Item 230.146. He chemical and volume control system components are located inside the containment which is a seismic Category I structure.

The chemical and volume control system can be isolated from the reactor coolant system in the event of a break or other adverse condition. The isolation valves between the reactor coolant system and chemical and volume control system are designed and qualified for design conditions that include D# 230.145-1

FSER Open item

~

closing against blowdown flow with full system differential pressure. These valves are qualified for adverse seismic and environmcntal conditions. The valves are subject to inservice testing including 1.

operability testing. He operability testing is a diagnostic test against operating flow and differential pressure..

The chemical and volume control system insitic containment is has a design pressure of 3100 psig and would not be expected to fail at reactor system operating pressure, his pressure exceeds the reactor coolant system design pressure. Therefore, the chemical and volume control system purification loop is not subject to an intersystem LOCA due to overpressure.

Chemical and volume control system leakage inside containment is detectable by the reactor coolant leak detection function as potential reactor coolant pressure boundary leakage. This le kage must be identified before the reactor coolant leak limit is reached. He nonsafety related classification of the system does not impact the need to identify the source of a leak inside containment.

10CFR50.55a(c) requires the following:

Reactor coolant pressure bour.dary. (1) Components which are part of the reactor coolant #

pressure boundary must meet the requirements for Class I components in Section III of the ASME Boiler and Pressure Vessel Code, except as provided in paragraphs (c)(2), (c)(3), and (c)(4) of this section.

L (2) Components which are connected to the reactor coolant system and are part of the reactor coolant pressure boundary as defined in Sec. 50.2 need not meet the requirements of paragraph (c)(1) of this section, Provided:

j (i) In the event of postulated failure of the component during normal reactor operation. -

l

- the reactor can be shut down and cooled Cown in an orderly manner, assuming makeup is provided by the reactor coolant makeup system; or (ii) The component is or can be isolated from the reactor coolant system by two valves in series (both closed, both open, or one closed and the other open). Each open valve must be capable of automatic actuation and, assuming the other valve is open, its closure time must be such that, in the event of postulated failure of the component

- during normal reactor orration, each valve remains operable and the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system only.

De purification loop of the CVS can be isolated by three safety related valves in series in both the letdown and return lines. He valves with motor and air operators mceive a signal to close in the event of law level in the pressurizer. Check valves are self actuated. These valves are included in the inservice testing requirements. The valves are designed and qualified design for conditions that include closing against full differential pressure.

De AP600 chemical and volume control system does not have a safety relate ! 'eiction to provide reactor coolant makeup. S*ty-related cort. makeup tanks are capable of provic% sufficient reactor coolant makeup for shutdown and cooldown without makeup supplied by the CVS. He safety-related means for safe shutdown of the reactor uses safety related systems and components including the passive core cooling system heat exchanger. See Section 7.4 for a discussion of safe shutdown of the reactor. If available, other nonsafety-related systems are also used for safe shutdown and cooldown of UD" 230.145-2

FSER Open Ibm ,. ,

the reactor wi'hout requiring use of CVS makeup.

SSAR Revision:

In Appendix 1 A add a discussion for criterion C.' n follows:

Crl*eria Referenced AP600 Section Criteria Position Clarification / Summary Descript'on of Exceptions Reg. Guide 1.26, Rev. 3,2n6 - Quality Group Classifications and Standards Dr Water , Steam , and Radioactive Waste Containment Components of Nuclear Power P%nts l C.I Exception A portion of the chemical and volume control I system that is defined as reactor coolant pressure I boundary uses an alternate classification in I conformance with the requirements of 10 CFR i 50.55a(a)(3). "Ihe alternate classification is I discussed in Section 5.2.

l In subsection 3.1.1. revise the fourth paragraph of the AP600 Complianee for Criterion 1 as follows:

Design, procurement, fabrication, inspection, and testing are performed according to recognized ccdes, standards, and design criteria that comply with the tequirements of 10 l- CFR 50.55a. As necessary, supplemental st mdards, design criteria, and requirements are i developed by the AP600 designers. A portion of the chemical and volume control system i that is defined as reactor coolant pressure boundary uses an alternate classification in I conformance with the requirements of 10 CFR 50.55a(a)(3). The alternate classification is I discussed in subsec: ion 5.2.1.3.

In subsection 3.1.2 add the following to the AP600 Compliance for Criterion 14 as fellows:

1 A portion of the chemical tnd volume ;ontrol system that is defined as reactor coolant I pressure boundary is nonsafety related. This portion of the system is capable of being I automaticahy isolated by safety-related valves that are designed and qualified for the i design requirements.

In subsection 3.1.4 revise the first paragraph under AP600 Compliance for Criterion 30 as follows.

Reactor coolant pessure boundary components are designed, fabricated, inspected, and I tested in conformance with the ASME Code,Section III. A portion of the chemical and I volume control system that is defined as reactor coolant pressure boundary uses an I alternate classification in conformance with the requirements of 10 CFR 50.55a(a)(3). The I alternate classification is discussed in Section 5.2.

In subsection 3.1.4 revise the first paragraph under AP600 Compliance for Criterion 31 as follows:

Control is maintained over material selection and fabrication for the reactor coolant i pressure boundary components so that the boundary behaves in a nonbrittle manner. The W25dfl M 8 230.145-3

FSER Open item _

g I ponion of the chemical and volume control system that uses an alternate classification is I not required to meet the requirements to prevent brittle failure. The reactor coolant pressure boundary materials exposed to the coolant are corrosion-resistant stainless steel or nickel <hromium-iron alloy. He nil-ductility transition reference temperature of the reactor vessel structural steel is established by Charpy V-r.otch and drop weight tests in accordance with 10 CFR 50, Appendix G (Reference 1). See Section 5.3 for additional information.

In subsection 3.1.4 revise the first paragraph under AP600 Compliance for Criterion 32 as follo'vs:

De design of the reactor coolant pressure boundary provides accessibility to the internal surfaces of the reactor vessel and most extemal zones of the vessel, including the nozzle-to-reactor coolant piping welds, the top and bottom heads, and external surfaces of the reactor coolant piping, except for the area of pipe within the primary shit.ld concrete. The inspection capability complements the leakage detection systems in assessing the integsity of the pressure boundary components. The reactor coolant pressure boundary will be periodically inspected under the provisions of the ASME Code,Section XI. Section 5.1 I provides the reactor coolant system primary loop drawings. De ponion of the chemical I and volume control system that uses an alternate classification is constructed to I requirements that do not require inservice inspection.

In subsection 3.2.2.6 add the following after the founh paragraph.

I A ponion of chemical and volume control system is defined as the reactor coolant pressure I boundary and is Class D ihis ponion of the chemical and volume control system is I seismically analyzed. See subsection 5.2.1.1 for the scismic analysis requirements.

Revise the third paragraph of Section 5.2 as follows:

The term reactor coolant system, as used in this section, is defined in Section 5.1. The l AP600 reactor coolant pressure boundary is consistent with that of 10 CFR 50.2. e*eept thaHhe-be=da y =ds 2: $e $ird incla:ica =he be:w:= i: :=c:ct ecc!=: sys::m =d de chemi=! =d eclum: centrc! sys::= Section 5.2.1.3 provide; i: ju; ifi=:ica for a 14r.i::d exemp:ica c $: defini:!ca cf 1: :=cter ecc!=: pr:=ure Sc=da y.

Revise the first paragraph of subsection 5.2.1.1 as follows:

Reactor coolant pressure boundary components are designed and fabricated in accordance I with the ASME Boiler and Pressure Vessel Code,Section III. A ponion of the chemical I and volu'me control system inside containment that is defined as reactor coolant pressure i boundary uses an attemate classification in conformance with the requirements of 10 CFR I 50.55a(a)(3). Systems other than the reactor coolant system connecting to the chemical I and volume control system have required isolation and are not classified as reactor coolant I pressure bourdary. The alternate classification is discussed in Section 5.2.1.3. He quality group classification for the reactor coolant pressure boundary components is identified in subsection 3.2.2. He quality group classification is used ta determine the appropriate sections cf the ASME Code or other standards to be applied to the components.

T Westinghouse 230.145-4

FSER Open item I 5.2.I.3 Alternate Classification E==ptice to 10 CFR R3 The Code of Federal Regulations, Section 10 CFR 50.55a requires the reactor coolant pressure boundary be class A (ASME Boiler and Pressure Vessel Code Section III, Class 1). Components which are connected to the reactor coolant pressure boundary that can be isolated from the reactor coolant system by two valves in series (both closed, both open, or one closed and the other open) with automatic actuation to close can be classified as class C (ASME Section III. class 3) according to 50.55a.

I A portion of the enemical and volume control system inside containment is not classified I as safety related. He classification of the AP600 definitica of reactor coolant pressure I boundary deviates from the requirement that the reactor coolant pressure boundary be I classified as safety related and be constructed using the ASME Code,Section III as I &firitica provided in 10 CFR 50.55a. 2 fe: $: ==:cr ecc!=: p==ux Sc=drj in that i 1: ==:cr ecc!=: p=== be=drj ec==:ic= :c 1: :hemied =d =!um: =nec!

l sy := pu-ifi=tica !cep n& i=ik an: dam =: The safety-related classification of the i AP600 reactor coolant pressure boundary ends at the third isolation valve between the I resctor coolant system and the chemical and volume control system, ne nonsafety-related I portion of the chemical and volume control system inside containment provides i purification of the reactor coolant and includes heat exchangers, dimineralizers, filters and I connecting piping. For a description of the chemical and volume control system refer to I subsection 9.3.6. He portion of the chemical and volume control system between the I inside and outside containment isolation valves is classified as Class B and is constructed i using the ASME Code,Section III. He juniS=:ica fc: $i: =:mp:ica :: b= d en $:

diff:==: be:==n $c AP600 =d p!=:: in =: =d =&: ==i&=:i= a: $: :im: $:

&5t= fer ==:c ecc!=:-pressure-bc=drj = "-i::=

1 The nonsafety-related portion of the chemical and volume control system is designed using i ANSI B31.1 and ASME Code,Section VIII for the construction of the piping, valves, and I components. The nonsafety related portion of the CVS inside containment is analyzed I seismically. The methods and criteria used for the seismic analysis are similar to those I used of seismic Category 11 pipe and are defined in the subsection 5.2.1.1. The chemical 1- - and volume control system ccmponents are located inside the containment which is a I seismic Category I structure.

I The alternate classification of the !!ri::d ==9ti= f cm 1: &Eni:i= cf ==:c ecc!=:

==

= bc=drj =d $: 6:ign cf 1: AP6" vi$ c nonsafety-related purification subsystem satisfies the purpose of 10 CR 50.2 =d 10 CFR 50.55a =d $ &finiti= cf

==:= =c!=: p== = bc=drj that structures, systems, and components of nuclear power plants which are important to safety be desi l;ned, fabricated, :rected, and tested to I quality standards that reflect the importance of the safety functions to be performed.

I I ne AP600 chemical and volume control system is not required to perform safety-related I functions such as emergency boration or reactor coolant makeup. Safety-related core I makeup tanks are capable of providing sufficient reactor coolant makeup for shutdown and I cooldown without makeup supplied by the chemical and volume control system. Safe I shutdown of the reactor does not require use of the chemical and volume control system

[ W8Stiligh00$e 230.145-5

FSER Open item I

makeup. AP600 safe shutdown is discussed in Section 7.4. W =: cf acn=f::y wh::d ch:mbc! =d vc!;me con:re! ;y;::= de:: nc: ==l: in : = duction of =f::y =d "! ==k 1 in adcetica cf ececpien ! mdictica =pe== =d !:n ;;=:m:i= cf md!= :iv: ==::.

I I The isoletion valves between the reactor coolant systt m and the chemical and volume control system are active safety-related valves that are designed, qualified, inspected and I tested for the isolation requirements. The isolation valves between the reactor coolant I system and chemical and volume control system are designed aad qualified for design I conditions that include closing against blowdown flow with full system differential I pressure, nese valves are qualified for adverse seistuic and environmental conditions.

I The valves are subject to inservice testing including operability testing.

I The potential for release of activity from a break or leak in the chemical and volume I control synem is minimized by the location of the purification subsystem inside I containment and the design and test of the isolation valves. Chemical and volume control I system leakage inside containment is detectable by the reactor coolant leak detection Y I function as potential reactor coolant pressure bounJary leakage. This leakage must be I identified before the reactor coolant leak limit is reached. The nonsafety-related I classification of the system does not impact the need to identify the source of a leak inside

i containment.

l-Revise the first paragraph of subsection 5.2.3.1 as follows.

i Table 5.2-1 lists material specifications used for the principal pressure-retaining

} applications in Class 1 primary components and reactor coolant system piping. Material specifications with grades, classes or types are included for the reactor vessel components, steam generator components, reactor coolant pump, pressurizer, core makeup tank, and the passive residual heat removal heat exchanger. Table 5.21 lists the application of nickel-chromium-iron alloys in the reactor coolant pressure boundary. He use of nickel-chromium-iron alloy in the reactor coolant pressure boundary is limited to Alloy 690.

. Alloy 600 may be used for cladding or buttering. Steam generator tubes use Alloy 690 in the thermally treated form. Nickel-chromium-iron alloys are used where corrosion resistance of the alloy i. an important consideration ar.d where the use of nickel-chromium iron alloy is the choice because of the coefficient of thermal expansion.

Subsection 5.4.3 defines reactor coolant piping. See subsection 4.5.2 for material specifications used for the core support structures and reactor intemals. See appropriate sections for intemals of other components. Engineered safeguards features materials a:e I included in subsection 6.1.1. He nonsafety-related portion of the chemical and volume

i. control system inside containment is constructed of corrosion resistant material that is I compatible with the reactor coolant pressure boundary. The nonsafety related portion of the I chemical and volume control system is not required to conform the process to I requirements outlined below.

3 M@lse 230.145-6

- PSER Open item ..

Revise subsection 9.3.6.1.1 as follows:

9.3.6.1.1 Safety Design Basis

- Re safety functions proviued by the chemical and volume control system are limited to containment isolation of chemical and volume control system lines penetrating containment, termination of inadvertent reactor coolant system bcron dilution, bolation of makeup on a steam generator or pressurizer high level signal, and preservation of the reactor coolant system pressure boundary, including isolation of normal chemical and I volume control system letdown from the reactor coolant system. As outiined in Section I_ 5.2, a ponion of the chemical and volume control system inside conNnment that is I defined as reactor coolant pressure boundary is nonsafety related and uses an alternate I classification.

Revise subsection 9.3.6.3 as follows:

9.3.6.3- Component Descriptions ne general descriptions and summaries of the chemical and volume control system components are provided below. The key equipment parameters for the chemical and l- volume control system components are contained in Table 9.3.6-2. Information regarding I component classifications is available in Section 3.2. The purification subsystem inside I containment is defined as reactor coolant pressure boundary. This subsystem is nonsafety-I related and is constmeted using standards outlined in subsection 3.2.2.6 to the ASME I Code Section 111. See Section 5.2 for additional information on analysis requirements. >

f MM 230.145-7 1

f i

A

L FSER Open item nu

. . ==

m 1

g 250.35F (OITS #6521)

FSER open item 230.147F, related to the CVS classification issue, requested that Westinghouse perform ASME Section XI leak testing on the isolation valves between the RCS and the CVS piping.

Although these valves do not strictly meet the definition of PIVs (two normally closed valves between high pressure and low pressure systems), the staff believes that the function of these valves is equivalent to PlVs for isolating the RCS from the non-ASME code class CVS piping. The potential for full RCS differential pressure across the valves from a failure of the downstream CVS piping dictates that leak tightness be assured so that failure of the CVS piping will not result in a LOCA for the AP600. Westinghouse should include valves CVS PL- V001, V002, V003, V080, V081, V082, V084, and V085 as PIVs under proposed SSAR Table 3.9-18 and subject them to the RCS PIV Integrity technical specification. TS 3.4.16 Limiting Conditions of Operation. In addition, these valves should be included in IST Table 3.916 as Category A valves requiring a leak test.

Response

The reasons for leak testing of these valves are different from the reasons for leak testing the pressure isolation valves (PlVs) included in the tech specs. The leak test requirement for the CVS valves should not be included with the tech spec requirements for PIVs. As discussed with the NRC staff, the requirement for leak testing of the valves at the interface between the CVS and RCS will be included in the inservice test plan included in SSAR subsection 3.9.6 and Table 3.916.

The discussion on the requirements for testing of the CVS. valves and the associated SSAR changes are included in the response to FSER Open Item 230.147.

SSAR Revision: NONE d

T westinghouse no.3.i

', FSER Open item ,,,

250.37F (OITS #6523)

The BASES background discussion of TS 3.4.16 should be revised to explain that the normally open

RCS to CVS interface valves will be treated as PlVs for the AP600 design because of the potential of '

these valves to perform a PlV functions if the CVS piping were to experience a failure. Westinghouse should develop appropriate wording for a PlV inclusion criterion explaining that the NRC rationale for designating the RCS to CVS interface valves as PlVs is based on the capability of these valves to establish and maintain the leak tight integrity of ASME Section 111 reactor coolant pressure boundary

- piping under high pressure RCS operation. Westinghouse should review the entire TS 3.4.16 BASES discussion and make it consistent with inclusion of the RCS to CVS interface valves.-

Response

The leak test requirement for the CVS valves will not be included with the Technical Specifications or bases. The leak test is included in the inservice testing. See the response to FSER Open items 250.35 and 230.147 for a discussion of the requirements for leak testing of these val,es.

SSAR Revision: NONE e

INW 250.37-1

& U

NRC FSER OPEN ITEM jpy Question 410.3?6F (OITS 6101)

Re:

NUREG/CR 0660 recommends the use of a three way thermostat temperature control valve for directing the engine water to the bypass or cooler. It should be of or equivalent to the "Amot" brand with an expanding wax type, temperature sensitive element, in its response to RAI Q410.183 dated August 3,1994, Westinghouse states that the SDECS is designed to use three-way thermostat water temperature control valves. Dese valves split the water fiow between the cooling radiator circuit and the engine return inlet circuit such that the engine cooling inlet circuit temperatures remain naarly constant under various eng.'ne loads and ambient temperature conditions. However, Westinghouse needs to verify that the ti.ree way thermostat has an expanding wax type temperature sensitive element and add this information to the SSAR. (OITS No. 333).

l Response: (Revision 1) l The motor-operated temperature control valve is replaced with a self contained temperature control valve. Based

! r- rr. :e 3 $ ^.me: Cant:P % 'ue: Ca 'ag. the phrase "enpanding wa- :ype" :' m: ; proper decerip::c of l the:enper ::ureentn4-valveHhereferedhi+phr,r =$not k =ed. The SS AR revisions below reflect this change.

I SSAR Revision:

l The fifth paragraph of SSAR subsection 8.3.1.1.2.1 Revision 19 will be revised as shown below:

De diesel-generator engine cooling system is an independent closed loop cooling system, rejecting engine heat through two separate roof-mounted, fan cooled radiator.. The system consists of two separate cooling loops ecch maintained at a temperature required for optimum engine performance by separate engine-driven coolant water circulating pumps. One circuit cools the engine cylinder bhick. Jacket, and head area, while the other circuit cools the oil cooler and turbocharger aftercooler.

He cooling water in each loop passes through a three-way self-contained temperature control valve which modulates the flow of water through or around the radiator, as necessary, to maintain required

\ water temperature. The tenverature control ralve has an e.tpanding wantype temperature sensitive l clenent or equiralent. The cooling circuit, which cools the engine cylinder blocks, jacket, and head areas, includes a keep-warm circuit consisting of a temperature controlled electric heater and an ac motor-driven water circulating pump.

I SSAR Figure M.3.1 4 N re ::.eJ = Ar :he 'r"^ "ng pages: in Revision 19 to shows this change.

4 W 410.326F(R1)-1

- Westle use 1-

-