NRC-94-4181, Submits Responses to NRC Requests for Addl Info Re AP600
| ML20070D207 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 06/27/1994 |
| From: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NTD-NRC-94-4181, NUDOCS 9407070237 | |
| Download: ML20070D207 (300) | |
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Westinghouse Energy Systei.is gagp Electric Corporation NTD-NRC-94-4181 DCP/NRC0117 Docket No.: STN-52-003 June 27,1994 Do ument Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION:
R.W.BORCHARDT SUBJECr:
WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION ON THE AP600
Dear Mr. Borchardt:
Enclosed are three copies of the Wes'.inghouse responses to NRC requests for additional information
' the AP600 from your letters of March 16,1994, April 7,1994, April 19,1994, April 29,1994, May 2,1994, May 11,1994, May 12,1994, May 16,1994 May 18,1994, May 19,1994, May 23, 1994 and May 26,1994. In addition, revisions of responses previously submitted are providc<i.
A listing of 11.
IRC requests for additional information responded to in this letter is contained in Attachment A.
These responses are also provided as electronic files in Wordperfect 5.1 fonnat with Mr. Kenyon's copy.
If you have any questions on this material, please contact Mr. Brian A. McIntyre at 412-374-4334.
h Nicholas J. Liparulo, Manager Nuclear Safety Regulatory And Licensing Activities
/nja Enclosure cc:
B. A. McIntyre - Westinghouse T. Kenyon - NRR (9.<'
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- NTD-NRC-94-4181 ATTACIIMENT A AP600 RAI RESPONSES SUBMITTED JUNE 27,1994 RAI No.-
Issue 210.009R01i Piping support in modules 210.029 i
SSAR Section 3.2.2.5 210.031 l
Requested revision to RAI 210.9 210.034 SSAR Table 3.2-3 21.336 i
Component cooling water quality group 210.038 PRHR support safety class 210.039 l
Class A PRHR HX/ Class C IR'VST interface 210.040 l
SSAR Section 3.6.2.1.1.4 210.043 l
SS AR section 3.6.2.3.4.2 210.047 SSAR section 3.7.3.4 210.050 i
SSAR section 3.7.3.8.2.1 210.053 l
SS AR section 3.9.2.1 210.054 i
SSAR section 3.9.2.1.2 210.055 SSAR sections 3.9.2.1.2 & 14.2.8.2.18 210.056 SSAR section 3.9.2.1
-210.057 i. SS AR sections 14.2, 1.9.2.1.1, 3.9.2.1.2 210.058 l
SSAR section 14.2.8.1.77 -
210.072 l
SSAR sections 3.9.7.1 & 3.9.7.3 '
210.082 Dynamic kads for electrical equipment 210.083 Superimposition of accident load onto seismic load 210.086--
l Auditable records - seismic / dynamic qualification
. 210.092 i
Determination of worst case orientation l
- 210.093 l
Aging by analysis l
210.094 Functionality of CRDM for seismic & LOCA loads 210.096 SSAR section 3.9.5 i.
210.099 l ' SSAR figures 3.9-5 & 3.9-6 l
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i NTD-NRC-94-4181 ATTACIIMENT A AP600 RAI RESPONSES SUBMITTED JUNE 27,1994 RAI No.
Issue 210.101 Kev dimensions of reactor vessel & supports 210.106 ASME fatigue design curve marcin for 60-vr plant 220.061R0l!
Waterticht/airticht seal in SSAR 3.8.2.1.1 220.091 Eauipment hatch stmetural calculations 230.011R01:
Seismic 11 building structures 230.054 Zone 3 reauiremt....s of UBC for analysis 230.058R01:
flich frequency modes of structures 230.063R0l!
Soil column properties for horiz. & vert. models 230.066 Adeauacy of Zone 2A reauirements of UBC 230.068 Use of concentric or dual zone systems i
230.068R01:
Use of concentric or dual zone systems i
230.073 Clarification of SSAR Section 3.2.1.1.2 260.022 Use of external organizations 270.004 Eauipment list reso 270.005 Compliance with EO reauirments 410.130 Failure of TB CCWS i
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410.191 Containment of flood waters
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'410.193 Protection of battery rooms from spray / flooding 410.205 Moderate-energy systems 410.207 Exclusion ofcomponents as missile sources 410.211 Inclusion of references in SSAR 410.212 Safety-related cauipment in TG strike zone 410.215 Protection by distance 410.217 Credible missile sources 410.218 Methodology to determine missile containment 410.220 Credible missile source inside containment 2
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NTD-NRC-94-4181 ATTACHhfENT A AP600 RAI RESPONSES SUBhilTTED JUNE 27,1994 RAINo.
Issue 410.222 Fans inside containment as credible missile source 410.224 Barries inside containment as missile shields hiissiles from nor.s$fety SSCs inside cont 410.225 1
410.226 hiissile protection by distance inside cont 410.227 MCR erotection from missiles inside cont 410.228 Systems needing protection 410.229 Probability of wind speeds arcater than 300 mph 410.230 h1CR missile protection 410.231 Use of 2 psi pressure droo 440.075 Sizing of pressurizer safety valves 440.077 Environmental conditions for PZR safety valves 440.082 TS LCO 3.4.13. increasinu RCS temperature 440.084 Ouality croup classification for LTOP system 440.086 Availability of PXS durina refuelina conditions 440.089 ADS actuation freauency 440.091 Break in PRHR heat exchancer common lines 440.099 SSAR Table 6.3-1 sheet 1 440.102 Accumulator P&lD -
440.107 SSAR Section 6.3.3.1.1 440.114 Loss of NRHRS during refuelina 440.115 ADS actuation afler extended loss of ac power 440.127 RNS safety classification 440.129 RNS cooling with IRWST 440.135 Pathway from IRWST to refueling cavity via RNS 440.139 Inadvertent closure of RNS suction isolation valve 440.157 Controls on isolation of unborated water sources 3
NTD-NRC-94-4181 ATTACHMENT A AP600 RAI RFSPONSES SUBMITTED J).NE 27,1994 RAINo.
Issue 440.159 Desien of pressurizer safety valves 440.161 Remote ventine of PRHR heat exchancers 460.019 Incorporation of RAI responses in SSAR 460.023 Cont. filtration system compliance with RG 1.140 480.062 Steam nenerator isolation valve closure time 480.072 Credit taken for secondary containment durine DBA 490.004 Use of emperical data 490.006 Comparison of nozzle ETC with previous desien 490.007 Fuel performance evaluation control 400.008
[ Rod bow model 490 009 Definition of material corrosion rates 490.010 Justification for no waterloccine 491.001 Reactivity coefficients 491.002 Compliance conditions 491.003 Shutdown marcin uncertainties 4Pl.004 Skewed flux distribution 531.n05 Rod insertion limit calculation 491.006 t
Boron calculation 492.001 Thermal desien procedure parameters 492.003
. Rod bow penalty 492.004 W-3 DNB correlation 952.078 SPES-2 lower plenum conditions 952.083 SPES-2 pipine system bend / elbow radii 952.085 SPES-2 pressure vessel connections 952.087 SPES-2 upper plenum & annular downcomer 4
NRC REQUEST FOR ADDITIONAL INFORMATION iii"
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Response Revision 1 Question 210.9 Section 3.7.3.S.; ol the SS AR. " Piping Systems on hhnlules," states that umlules are construt ted using a structural steel tramework to support the equipment. pipe, and pipe supports in the module and that, with one excepion the framewoik is de3igned as part of the bdding structure. It'. su%equent le installation of the modules. the tramew oik is relied upon to support any portion of tne piping prmide the h or not complying with the jurishetional boundary rules in Section NF of Section ill of the ASME Code.
j Response (Revision 1)
The ASME Code, Section Ill. Subsection NF, does not clearly ad. bess junsthetional boundaries hir the specilic case of nmlule structural steel tramework. Th module tramework u consiocred to be a part of the buihling structme th.a is built in the shop hir cornenience. The inodule steel f rame pertonns multiple functions (such as supporting maintenance plationus, utilities, lightmg. shiehling, etc.) and supports the c omponents and piping. Figure 210.09-1 illustrates a large module and shows the NF jurisdictional boundaries. Figure 210.04-2 illustrates a smaller module and similarly show s jurisdictional boundaries. The module framework is detaded on mislule steel drawings rather than on indnidual pipe support or cotoponent support drawings.
Modules comaining salety-related equipment are classified as seismic Categ iry I. The module steel f rames are designed and constructed to ANSI!AISC N690, as described in SS AR Subsectio i 3.8.4. The module trame structural quablication includes the loads applied to it by the pipe and component supp rts, in addition to the loads exerted by the maintenance plationns, utilities, ete Structural steel extending f rom th: unlule framework to the pipe or component (or standard component support) complies with Subsection NF of 1ie ASME Code.
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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 g
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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 NF P!DF 'luio()RT
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NRC REQUEST FOR ADDITIONAL INFORMATION Question 210.29 With respect to quality group classification of certain sy stems, components, and equipment, the staff does not completely agree with the information in Section 3.2.2.5 of the SSAR, " Equipment Class C;" Appendix 1 A of the SSAR, " Conform 2mee with Regulatory Guides;" the response to Q 210.1 (dated December 22, 1992); and the exceptions to Section 3.2.2 of the Standard Review Plan (SRP) in Revision I of WCAP-13054.
Section 3.2.2.5 of the SSAR states that items that perform one or more of the following safety-related functions are classified as Class C (Quality Group (QG) C):
Provide safety-injection or maintain sufficierit reactor coolant insentory to allow for core cooling Provide core cooling Pros ide containment cooling Provide for removal of radiation from the containment atmosphere as necessary to meet the offsite dose limits To be consistent with Regulatory Guide fRG) 1.26 and ANSI /ANS 51.1, the staff's position is that all items that perform one or more of the above functions should be Class B (t)G B).
In Appendix 1 A of the SSAR and WCAP-13054, exception is taken to Position C.I.a of RG 1.26, which is the basis for the staff position stated abose relative to Section 3.2.2.5. The basis stateJ in the SSAR for this exception is that for the AP600. QG B is resersed for the containment boundary including the containment isolation salves. The exception acknowledges that for QG C, the ASME examination and inservice inspection rules are less stringent than l
those for QG B.
The SSAR states that QG C is acceptable for pasrive safety-systems components such as the l
accumulators and the IRWST, and that minor leakage is not a problem because (a) these components are inside l
containment (b) minor leakage does not affect the component's functional performance, and (c) there is continuous l
water level and gas pressure monitoring of the accumulators that detects leaks. The following is the staff's position relative to this exception:
As stated in Section 6.3 of the SSAR, the primary function of the passive core cooling system is to provide emergency core cooling following postulated design basis events. RG 1.26 classifies emergency core cooling systems as QG B. The system boundary includes those portions of the system required to accomplish the specified safety function. and connee:ed piping up to and including the first valve that is either closed or capable of automatic closure when the safety function is required. This is irrespective of the fact that the system does not recirculate post-accident fluid outside containment. In addition. the position stated in the SSAR for the exception relative to the ability of the accumulators and the IRWST to accommodate minor leakage appears to be similar to the piping leak-before-break (l.BB) issue, but without a technical basis to implement LBB for these components. The staff cannot accept such an argument as the basis for the design of an emergency core cooling system. Therefore, the staff's position is that these components should be QG B.
210,29-1 W Westinghouse
r NRC REQUEST FOR ADDITIONAL INFORMATION A
The response to Q210.1 contains positions similar to those in Section 3.2.2.5 and Appendix 1 A of the SSAR, and is not acceptable for the same reasons discussed above. Therefore, revise the applicable information in Appendix 1 A relative to RG 1.29, Section 3.2 2.5, Table 3.2-3, Section 6.3 (including the P&lDs), WCAP-13054, and the response to Q210.1 to comply with the above staff positions.
Response
Smee the AP600 SSAR was submitted, ANS has deseloped standard ANS-58.14, Safety and Pressure Integrity Classification Criteria for I.ight Water Reactors. It has been review ed and all of the Nuclear Pow er Plant Standards Committee members have approved this standard, including the NRC member. The standard is in the process of being issued. This ANS standard contains criteria that are very similar to the AP600 equipment classification criteria. This standard separates safety classification and the pressure integrity classification.
The safety classi6 cation in this standard is either safety-related, supplemental, or nonsafety-related. The safet3-related classification is not subdivided based on function. Safety-related equipment is required to:
linsure the integrity of the reactor coolant system pressure boundary, or Ensure the capability to shut down the reactor and maintain it in a safe shutdown condition, or Ensure the capability to prevent or mitigate the consequences of accidents that could result in off-site expsures in excess of NRC limits The pressure integrity classification in this standard is separately assigned. It identifies w hich items will be subject to the requirements of the ASME Boiler and Pressure Vessel Code, Section 111 or other codes. There are 5 pressure integrity classifications. ANS has developed criteria to assign these pressure integ.-ity classifications. These criteria assign C-1 to those items that form the radioactive material barrier of highest importance (i.e., closest to the reactor core) and C-2, -3, -4 are assigned successisely to items that form less important radioactive barriers. The ASME Code, Section 111, part NB NC, ND apply to the top three pressure integrity classes respectisely. ANS 58.14 uses specific criteria to assign these pressure integrity criteria. These criteria are:
Class C-1: Maintain pressure integrity of the reactor coolant system pressure boundary Class C-2: Maintain pressure integrity of
- The containment boundary and its extensions
- Safety-related systems (emergency core cooling, emergency containment cooling, emergene) containment radioactivity removal, emergency decay heat removal) that are located outside containment AND that communicate with the reactor coolant system pressure boundary or the containment interior AND that are not isolated following design basis accidents.
Class C-3: Maintain pressure integrity of other safety-related equipment This ANS standard results in safety injection accumulators being safety-related since they mitigate design basis accidents. Accumulators would be pressure integrity class C-3 since they are not part of the reactor coolant system pressure boundary (C-l) nor are they part of the containment boundary (C-2) and they are safety-related equipment that is not located outside containment. Applying ANS standard 58.14 to the IRWST and the passive containment water storage tank results in them being classified as pressure integrity class C-3.
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NRC REQUEST FOR ADDITIONAL INFORMATION g)
Westinghouse thinks that ANS standard 58.14 represents a comprehensive safety classification system that has been developed with broad industry input and consensus. The Westinghouse APM)0 equipment classification catagories (A,11. C) are consistant with the ANS 58.14 pressure catagories (C-1, C-2, C-3). Westinghouse therefore belieses that the APMX) equipment classification approach is appropriate.
SSAR Revision: NONE 210.29-3 W westingt100S8
NRC REQUEST FOR ADDITIONAL INFORMATION gm
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Question 210.31 The response to Q210.9 dated January 22,1993 relatis e to the jurisdicti~.al boundary betw een module framew orks and piping supports within the module appears to be acceptable. 1. - -
er, the staff does not agree with the 1"t paragraph in this response, which states that subsequent to the incoqt.ation of AISC N690 into the ASME Code, the design criteria for linear supports would change from Subsection NF to AISC N690. Subsection NF includes rules for construction of such supports, where " construction" is as defined in Subsection NF-1100(a).
Therefore, even after an acceptable incorporation of AISC N6W into the Code, the stalTs position is that Subsection NF, not AISC N690. will remain the only staff-endorsed rules for these supports. Revise the response
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to Q210.9 3 delete the last paragraph.
Response
The response to RAI 210.9 has been revised to delete the last paragraph.
1 SSAR Revision: NONE 0
l NRC REOUEST FOR ADDITIONAL INFORMATION Question 210.34 Table 3.2-3 of the SSAR, "APMK) Classineation of Systems and Components does not appear to include the classifications for the piping and supports of each system. Although these classifications may be implicit in this table, the staff's position is that this important table should explicitly include the classifications, principle constniction codes, and quality assurance programs for all piping and all supports for piping and equipment in each sy stem. The supports should have the same classification as the supported component or equipment. This information should be consistent uith applicable piping and instrumentation diagrams in the SSAR. Revise Table 3.2-3 to include this information.
Response
The classification for piping can be determined from the P&lDs. Each line has a three letter code to specify the material, pressure classilication, and equipment classification. The third letter corresponds to the APMK) equipment classification. This specification code is shown on Sheet 2 of Figure 1.7-2.
The classification of supports for piping and components is determmed by the classification of the component or piping supported. Suppcrts for APNX) equipment Class A, B, and C components and piping are constructed to ASNIE Code, Section 111, Subsection NF requirements. Requirements for supports for nonsafety-related component and piping are established by the principle construction code for the supported component or piping. For example, the requirements for supports for ASME-H31.1 piping are specified in ASME-lu t. l. The SSAR revision shown below explicitly states the requirements for supports.
SSAR Revision:
l Revise the first paragraph of Subsection 3.2A as follows:
The applicatior. of the APMx) equipment and seismic cNiicatio' system to APMX) systems and components is show a in Table 3.2-3. Table 3.2-3 lists mechanical and fluid system component and its associated equipment ele,s and seismic category as well as other related information. Valves a,J dampers w hich are classes A, B, C, D or R are listed. Vahes and dampers of other classes are generahy not listed. Supports for piping and components have the same classification the component or piping supported. Supports for AP600 equipment Class A, B, and C components and piping are constructed to ASME Code,Section III, Subsection NF requirements. The principle construction code for supports for nonsafety-related components and piping is the same as that for the supported component or piping, l
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NRC REQUEST FOR ADDITIONAL INFORMATION E n--)
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Question 210.36 in Table 3.2-3, Sheets 5 through 8, and Figum 9.2.2-2, Sheet 2, of the SSAR, the equipment and piping in the Component Cooling Water System (CC") is all classified as QG D, with the exception of the containment penetration area. Section 9.2.2 of the SSAR states that the CCS serses no safety-related function except for containment isolation. How ever, Table 9.2.2-2 of the SSAR lists the follow mg safety-related components for w hich
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the CCS prosides a reliable supply of cooling water:
Reactor coolant pumps (ASN1E Class 1)
Chemical Volume and Control System letdow n heat exchangers ( ASN1E Class 3)
Normal RHR heat exchangers and pumps ( ASNIE Class 3)
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Position 2b of RG 1.26 states that portions of cooling water sy stems that perform functions similar to those abose should be classified as QG C.
This means that the design of these components should meet all of the rules of l
ASN1E Class 3. Either revise Table 3.2-3, Section 9.2.:', Figure 4.2.2-2, and applicable portions of Appendix l A of the SSAR and Revision I to WCAP 13054 to reflect this staff position. or provide a detailedjustification for the AP600 position on this issue.
Response
The component cooling water system provides cooling water to the safety-related components identified above to support their normal operation. None of these safety-related components require cooling water to perform their safety-related functions. The safety-related functions of these components are limitca to maintaining primary coolant system integrity and for the reactor coolant pump to provide for flow coastdown of the reactor coolant. The safety-related function of transferring heat to the ultimate heat sink is provided by the passise safety systems by transferring heat through the containment shell to the surrounding air. hiaintenance of primary system integrity is independent of the cooling water supply to the safety-related component. Cooling water is not required for flow coastdown by the canned-motor reactor coolant pump. See Subsection 5.4. l.3.4 for a discussion of the coastdow n capability of the reactor coolant pumps. Accordingly, the component cooling water system is properly assigned a nonsafety-related classification.
See the response to R AI 410.117 for additional information.
SSAR Revision: NONE l
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4 NRC REQUEST FOR ADDITIONAL If4 FORMATION m
b Question 210.38 Section 5.4.14 of the SSAR stater that the passive residual heat removal heat exchanger (PRHR HX) is AP600 Equipment Class A and its supports are Class C. These supports are not listed in Table 3.2-3 and are not included in Figure 6.3-2 of the SSAR, "P&lD for Passise Core Cooling System." Figure 5.4-9 provides a sketch of these supports, but there is not enough detail to understand the support configuration. The staff's position is that supports for safety-related components and equipment should be the same safety class as the supported item. Revise the section, table, and figures to classify the supports for the PRHR HX as Equipment Class A, seismic Category 1.
Response
The passive residual heat removal heat exchange 4 is designed and fabricated according to the rul& of ASME code, Section Ill, class 1. This includes the pressure boundary parts of the heat exchanger and the supports that fall under the jurisdiction of ASME Code, Section 111, Subsection NF. Therefore, the supports for the passive residual heat removal heat exchanger are reclassified as Class A.
The supports are part of the passive residual heat removal heat exchanger, so they are not separately classified in Table 3.2-3 of the SSAR,nor shown on the P&lD.
SSAR Revision:
Paragraphs 4, 5 and 6 of Section 5.4.14.1 of SSAR will be revised as follow s:
The passive residual heat removal heat exchanger is designed and fabricated according to the ASME Code, j
Section 111, as a Class I component. Hose portions of the passive residual heat exchanger that support twru44he rwietor-the-rowtor--cookmt-twe+iu+-twwwi+wy primary-side pressure boundary and falls under the jurisJietion of ASME Code, Section 111, Subsection NF are AP600 equipment Class A ( ANS Safety Class 1 Quality Group A).
j Stresses for ASME Code, Section ill equipment and supports are maintained within the limits of Section III of the l
Code. Section 5.2 provides ASME Code, Section ill and material requirements. Section 5.2.4 discusses in-service inspection.
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NRC REQUEST FOR ADDITIONAL INFORMATION wmq Question 210.39 Section 6.3.2.2.5 of the SSAR states that the PRHR HX inlet and outlet piping connects to inlet and outlet channel heads mounted through the In-containment Refueling Water Storage Tank (IRWST) wall with a tubesheet that is part of the tank wall. The PRHR HX is Class A, and the IRWST is Class C. Section 3.8.3.1.7 states that the east wall of the IRWST consists of structural steel modules, filled with concrete and forming, in part, the refueling cavity, steam generator compartment, and pressurizer compartment walls. It is not clear to the staff what the relationship is between the IRWST tank wall, tubesheets, channel heads, and the structural steel modules. Appendix 3A and Figure 6.3-2 of the SSAR do not appear to contain this information. Revise the SSAR to provide a complete description of this area. Include sufficient sketches and information to (a) describe the abose relationships, (b) identify the AP600 Equipment Safety Class of each item and clearly identify the location of the interface between the Class A PRHR HX and the Class C IRWST, and (c) describe the details of heat exchanger inlet and outlet piping pass 'hroughs in the modules, including a description of applicable analyses.
Response
(a) The attached figure 6.3-5 show s how the channel head and the tubesheet of the passive residual heat removal heat exchanger are connected to IRWST (in-containment refueling water storage tank) tank wall. The IRWST is a structural steel module, filled with concrete and form elements to which the heat exchanger is attached.
As shown in the figure the extended flange is also attached to the housing frame for the tube bundle, w hich is attached to the IRWST on the top anu bottom. The tube supports supporting the tube bundle are attached to the housing frame.
(b) The passive residual heat removal heat exchanger is designed and fabricated to the rules of ASN1E code, Section Ill, Class 1. This includes the pressure boundary parts (tubes, tubesheets, and channel heads) and the supports I
(housing frame, tubes supports, U-flange, extension flange) that fall under the jurisdiction of the ASN1E code, Section 111, Subsection NF. So the AP600 equipment classification for the heat exchanger is Class A. This is l
consistent with the response to RAI 210.38. The IRWST is classified as a building structure nd is designed per ANSI /AISC N690 code. The nP600 equipment classification for IRWST is Class C.
(c) There is no inlet or outle' piping pass-through in the module (IRWST wa:1). Please refer to the attached figure 6.3-5.
The SSAR statement that the tubesheet is part of the in-containment refueling water storage tank (IRWST) wall in section 6.3.2.2.5 is incorrect and is corrected as shown below.
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SSAR Revision:
Figure 6.3-5 will be added to the SSAR.
l The second paragraph of Subsection 6.3.2.2.5 of the SSAR will be revised as follows:
The heat exchanger inlet piping connects to an inlet channel head located near the outside top of tankvmomwel timmglshe '
di-wah+tulW*vuh*M -twa4-tl*wank-walt The inlet channel itend and tubesheet are l
l W Westinghouse l
r NRC REQUEST FOR ADDITIONAL INFORMATION attached to the tank wall via an estension flange which is also attached to the housing frame. The housing frame which is a part of the heat exchanger support is attached to the IRWST floor and top. The heat exchanpr supports are designed to ASME Code, Section 111. Subsection N1'. He extended flange is designed to accommodate thermal expansion. ne figure'6.3-5 illustrates the relationship between these parts and the boundaries of dessa code jurisdiction. The heat exchanger outlet piping is connected to the outlet channel head, which is vertically below the inlet channel head, near the tank bottom. The outlet channel head has an identical structural configuration to the inlet chant el head. Iloth channel head tubesheets are similar to the steam generator tubesheets and they have manways for inspection and maintenance access.
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NRC REQUEST FOR ADDITIONAL INFORMATION s.
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NRC REQUEST FOR ADDITIONAL INFORMATION
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Question 210.40 Section 3.6.2.1.1.4 of the SSAR, "High linergy Piping it. Containment Penetration Areas" should be changed as j
follows to be consistent with staff positions in Section 3.6.2 of the SRP:
Revise the third bullet to include a conunitment that when guard pipes are used in this area, the enclosed portion a.
of Duid system piping should not only be seamless, but should not contain circumferential welds unless specific access provisions are made in the guard pipe to permit inservice solumetric examination of these welds in accordance with the augmented inservice inspection provisions in the fourth bullet of this subsection. If appli-cable, inspection ports in the guard pipe should not be hicated in that portion of the guard pipe passing through a shield building annulus.
b.
The definition of break exclusion areas in the first paragraph and last three bullets of this subsection are not completely acceptable. The staff's position on this issue is that this area can start at the inboard side of the inside isolation valve but must end at the outboard side of the outside isolation vahe. Revise these portions of this subsection to be consistent with the staff position.
Response
The third bullet is revised to state that there are no circumferential or longitudinalwelds in the piping enclosed a.
within the guardpipe. Details of the arrangement are show n in SSAR Figures 3.8.2 4, b.
The staff's position in Section 3.6.2 of the Standard Review Plan is not consistent with practice approved by NRC staff on previous projects. The break esclusion area should evtend to the anchor or restraint adjacent to the isolation valve. The Standard Review Plan position requires consideration of a break occurring in the short l
length of piping between the anchor (or pipe rupture restraint) and the isolation valve.
l The auxiliary building anchors on the main steam and feedw ater piping are as close as practical to the isolation valves with a short section of piping separating the valve and anchor to permit space for the branch connection for the main steam isolation vahe bypass, as well as inservice inspection of the welds. This portion of the pipe, including the anchor, is ASN1E Code, Section 111, Class 3 as shown on the Piping and Instrumentation diagram in SSAR Figure 10.3.2-1. The anchor protects the valve from breaks postulated in the turbine building.
A break is not postulated in the short section of pipe at the weld to the valve nor at the weld to the anchor.
Design of protection features to permit operability of the vahe for such a postulated break adjacent to the valve is impractical. On previous projects. the short leqth of pipe between the isolation valve and the anchot has been designed to meet the stress limits of the brea t exclusion zone, as is proposed for the APMU, and this has been considered to be a part of the break exclusion zone.
SSAR Revision:
Revise third bullet in Subsection 3.6.2.1.1.4 as follows:
The number of circumferential piping welds is minimized by using pipe bends in place of welding elbows when practicable. There are no longitudinal piping welds in the break exclusion zone. Where guard pipes are used, W Westingt10use I
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NRC REQUEST FOR ADDITIONAL INFORMATION there are no r*cumferential or longitudinal welds in the piping enclosed within the guardpipe. Details of the arrangement are shown in Figure 3.8.24.
a 210.40-2 3 Westinghouse 1
i NRC REQUEST FOR ADDITIONAL INFORMATION A
Question 210.43 Section 3.6.2.3.4.2 of the SSAR states that if energy absorbing material is used in the design of pipe w hip restraints, the allowable deflection is 80% of the maximum crushable height at uniform crushable strength. In accordance with Section 3.6.2.Ill.2.a of the SRP, the staff's position is that the allowable capacity of crushable material shall be limited to 899 of its rated energy dissipating capacity as determined by dynamic testing at loaded rates within 150% of the specified design loading rate. The rated energy dissipating capacity shall be taken as not greater than the area under the load-deflection curve as illustrated in Figure 3.6.2-1 of Section 3.6.2 of the SRP. Revise Section 3.6 2.3.4.2 to be consistent with the staff's position.
Response
The MR will be resised to incorporate the SRP position.
SSAR Rt ision:
Revise the third paragraph of Subsection 3.6.2.3.4.2 as follows:
Energy-Absorbing Material - This type of restraint consists of a crushable, stainless steel, internally honey-comb-shaped element designed to yield plastically under impact of the wi.ipping pipe. A de. sign hot position gap is provided between the pipe and the energy-absorbing material to allow unimpeded pipe motion during seismic and thermal movements. Figure 3.6-2 shows a typical example of an energy-absorbing material restraint. The allowable capacity of crushable material shall be limited to 80% of its rated energy dissipating capacity as determined by dynamic testing, at loading rates within i 50% of the specified design loading rate. The rated energy dissipating capacity shall be taken as not greater than the area under the loaddeflection curve as illustrated in Figure 3.6.2-1 of NUREG 0800. Standard Review Plan, Section 3.6.2, Revision 2.
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NRC REQUEST FOR ADDITIONAL INFORMATION A
Question 210.47 Revise Section 3.7.3.1 of the SSAR, " Basis for Selection of Frequencies." to include a conunitment to the guidelines of Section 3.9.2.ll.2.c of the SRP, i.e., to avoid resonance, the fundamental frequencies of components and equipment should preferably be selected to be less that 1/2 or more than twice the dominant frequencies of the support structure.
Response
The SSAR will be revised to add the statement that resonant frequencies are avoided.
SSAR Revision:
The first paragraph of Subsection 3.7.3.4 of the SSAR will be resised as follows:
The effect of the building amplification on equipment and components is addressed by the tToor response spectra method or by a coupled anal) sis of the building and equipment. Certain components are designed for a natural frequency greater than 33 hertz. In those cases where it is practical to avoid resonance, the fundarnental frequencies of components and equipment are selected to be less that one-half or more than twice the domirutnt frequencies of the support structure.
I 210.47-1 W WB5tiflgh00S8
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NRC REQUEST FOR ADDITIONAL INFORMATION A
Question 210.50 The last paragraph in Section 3.7.3.8.11 of the SSAR states that when the st:pporting system for auxiliary (branch) pipe is equipment, the supported pipe can be decoupled from the suppc-ting equipment using the same criteria as when the supporting sy stem is a piping sy stem with the rtm pipe size replaced by the minimum dimension of the equipment. Describe how the minimum dimension (length, width, and height) of the equipment establishes an equivalent deccupiing criteria to that of a r m pipe.
Response
The decoupling criteria for run pipes is Lased ca the ratio of the run pipe diameter (or moment of inertia) to the aranch pipe diameter (or moment of inertia). His ratio is equisalent to a stiffness ratio w hen typical span lengths are used for the rtm pipes. In order to extend the decoupling to equipment, the equipment stiffness must be at least as large as the run pipe stiffness. Therefore, the equipment stiffness at the noule to pipe weld must satisfy the following:
A. 200El t
L' w here K = Equipment stiffness along each of three orthogonal directions E = Young's Modulus I
= Moment of inertia of equivalent rtm pipe that can be decoupled for branch pipe analysis L = Span length of equivalent rtm pipe based on Table NF-3611-1, ASME Code Section ill l
In addition, in order to preclude signi6 cant SSE inertial ampli6 cation, the displacement of the equipment nonle should be less than i inch in each of three coordinate directions. (See response to Question 210.49)
SSAR Revision:
The fourth paiagraph of Subsection 3.7.3.8.2.1 will be revised as follows:
When the supporting system is equipment, the supported pipe can be decoupled from the supporting equipment using the same criteria as when the supporting system is a piping system with the run pipe noannal-tst.+-we t!=phr 4*khrand4Wghe stiffness of the equip.nent. The equipment stiffness replaced by the m4samm-dime-w stiffness at the noule to pipe weld must satisfy the following:
A. 200El t
L' where K = Equipment stiffness along each of three orthogonal directions E = Young's Modulus W WestinEfl00S8
NRC REQUEST FOR ADDITIONAL INFORMATION I
= Moment of inertia of equivalent run pipe that can be decoupled for branch pipe analysis L = Span length of equivalent run pipe based on Table NF-3til1-1, ASMii Code, Section Ill l
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210.50-2
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NRC REQUEST FOR ADDITIONAL. INFORMATION A-e Question 210.53 e 3.4 2. I of the SSAR states that the preoperational piping vibration, thermal expansion, and dy namic effects Sec n
t~ t will be conducted only on the first AP600 plant because standardization of piping design eliminates the need to test the response of piping to transients in subsequent plants.
The discussions of these tests in Sections 1J.2.8.1.78,14.2.8.1.82,14.2.8.2.18, and 14.2.8.2.20 are also limited to the first plant only. This is an unaccepta >1e commitment. %e purpose of these tests is to confirm that the applicable piping systems, restraints.
compon ots, and suppo:u. have been adequately designed, fabricated, and installed to withstand flow-induced dyna.ae loadings under the steady-state and operational transient conditions and to confirm that normal thermal
.aotion is not restrained. The staff believes that one major cause of excessise vibration or excessive pipe movement can be attributed to improper support installation or loss of snubber functionality. Standardization of piping design will not proside assurance that such discrepancies do not exist. Therefore, the staff's position is that these preoperational tests are required to be conducted on all AP600 plants in accordance with the criteria discussed in Sections 3.9.2.1.1 and 3.9.2.1.2. Revise Sections 3.4.2.1,14. 2.8.1. 78,14.2.8.1.82.14.2.8.2.18, and 14.2.8. 2.20 l
to be consistent witn the above position.
Response
The SSAR will be revised to commit to the piping vibration and dynamic response testing oudined in the SSAR Subsections 14.2.8.1.78,14.2.3.1.82, and 14.2.8.2.20 for esery AP600 plant. The test program outlined in the SSAR for thermal expansion testing is for each plant, see Subsection 14.2.8.2.18.
SSAR Revision:
Delete the sixth paragraph of Subsection 3.9.2.1 as follows:
i E aniewing of p:;,ng fn-ih nden-daring :-
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Revise the heading for Subsection 14.2.8.1.78 as follows:
14.2.8.1.78 Steady-State Vibration Monitoring of afety-Related and High-Energy Piping (First-Pfant-Only)
Revise the heading for Subsection 14.2.8.1.82 as follow s:
14.2.8.1.82 Dynamic Response (Rest-Plant-Only)-
Revise the heading for Subsection 14.2.8.2.20 as follows:
14.2.8.2.20 Dynamic Response (Rest-Plant-Onlyt-210.53-1 W W85tlngh0US0 l
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NRC REQUEST FOR ADDITIONAL INFORMATION
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Question 210,54 Section 3.9.2.1.1 of the SSAR, " Piping Vibration Details," states that if system vibration is evidenced during initial operation, the maximum amplitudes are measured and ictated to alternating stress intensity levels based on the guidance of ANSI /ASN1E ON1, " Operation and Maintenance of Nuclear Power Plants " Part 3. Howeser, the scope of this OM Standard is more broad than the brief discussian in this section. This standard provides genend requirements for the assessment of vibration in all safety-related piping systems during preoperational and start-up testing. It includes steady state and transient vibration testing, acceptance criteria, and recommendations for corrective action w hen required. In addition, it provides guidance for the assessment of vibration levels of applicable piping sy stems during plant operation. For the preoperational piping vibration and dy namie effects tests on all AP600 plants, the staff's position is that a commitment to the full scope of ANSI /ASME OM, Part 3 should be prosided. If exceptions are taken to any portion of this standard, they should be clearly delineate ' in the SSAR and the bases for such exceptions should be provided. Resise Section 3.9.2.1.1 to reflect this staff position.
Response
The commitment to perform piping sibration te. sting and assessment.a accordance with ANSI /ASME OM, Part 3
~
will be added to the SSAR.
SSAR Revision:
Revise the first paragraph of Subsection 3.9.2.1.1 as follows:
Piping vibration loadings can be placed in two categories: transient induced sibrations and steady-state vibrations. The first i. a dynamic sy stem response to a transient, time-dependent forcing function, such as fast valve l
closure. The second is a constant vibration, usually flow-induced. Piping vibration testing and assessment is performed in accordance with ANSl!ASME OM, (Reference 2) Part 3.
210.54-1 W W351lnah00S8 u
l NRC REQUEST FOR ADDITIONAL INFORMATION Question 210.55 To be acceptable, the discussions in Sections 3.9.2.1.2 and 14.2.8.2.18 of the SSAR require a more specific commitment to the preoperational piping thermal expansion test program procedures. The staff's position is 'aat these tests should be conducted in accordance with the ASME ON1 Standard Part 7, " Requirements for Thermal Expansio, Testing of Nuclear Power Plant Piping Systems " his standard contains procedures to be used for the assessment of thermal expansion response and design verification of piping systems. Implementation of this standard ensores that the piping system is ready for testing and can expand and contract as required during all plant conditir ns.-
' verifies that (a) expected expansion can be accommodated by the piping system restraints, (b) movement. not obstructed by any unintentional restraints, and (c) response is within design tolerances. It also pnvides guidance for development of acceptance criteria, instrumentation, and measurement techniques. as well as corrective actions and methodologies foi reconciling moven.nts that differ from those specified by the acceptance criteria. Revise uions 3.9.2.1.2 ano 14.2.8.2.18 to provide a specific commitment that detailed test specifications q
for thermal ex wing of piping systems during preoperational and start-up testing will be in full accordance with ASME ON rd. Part 7.
Response
Detailed test specilications ior thermal expansion testing of piping systems during preoperational and start-up testing will be in accordance with ASME OM Standard, Part 7.
SSAR Revision:
Revise the first paragraph in Subsection.5.9.2.1.2 as follows:
The piping thermal expansion testing program verifies that the piping systems expand within acceptable limits dering heatup and cooldown. Also, this program verifies that the standard component supports (including spring
- h..igers snubbers, and struts) can accommodate the expansion of the piping within an acceptable range for required modes of operation. Detailed test specitications for thermal expansion testing of piping systems during preoperational and start-up testing will be in accordanca with ASME OM Standard, Part 7.
Revise the Performance Criteria in Subsection 14.2.8.1.67 as follows:
Performance Crite:ia The reactor coolant system has operated at full-flow conditions for a minimum of 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> with the reactor coolant temperature at or above 515 F for at icast one-half of the operating time.
The performance criteria for individual systems are a part of the individual test procedures seqtanced by this procedure.
Test specifications for _ thermal expansion testing of piping systems during preoperational and start-up tesi.g are in accordance with ASME OM Standard, Part 7.
SM W WestinEhouse
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NRC REQUEST FOR ADDITIONAL INFORMATION Resise the Perfoun:mce Criteria in Subsection 14.2.8.2.18 as follows:
Perfarmance Criteria l'or the components being tested, the following apply:
There is no evidence of blo: Ling of the thermal expansion of any piping or component, other than by installed supports, restraints, and hangers.
Spring h,nger movements must remain within l'ie hot and cold setpoints and supports must r.ot become fully retracted or extended.
Piping and components must return to their approximate baseline cold position.
Detailed test specifications for thernut expansion testing of piping systerns during preoperational and start-up testing are in accordance with ASME OM Standard, Part 7.
210.55-2 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMAYlON Question 210,56
'lhe second puragraph in Section 3.4.2.1 of the SSAR states that the preoperational piping sibration, thermal expension, and o3 namic eflects test programs will be conducted on ASN1II Class I,2, and 3, and other high c icrgy piping sy stems. It is the staf f's position that thes,e test programs should include safety-related instrinnent sensing lines at least up to the first support in these lines. Revise Section 3.4.2.1 to proside a commitment to include such lines in these test programs.
Response
These sensing lines will be included in the preoperational testing programs up to he first support in each of three orthogenal directions from the process pipe or equipment connection point.
SSAR Revision:
Resise the first paragraph of Subsection 3.4.2.1 as follow s:
A preoperational test program as described in Section 14.2 is implemented as required by Nil-3622.3, NC 3622, and NI)-3611 of the ASN11! Code Section ill to verify that the piping and piping restraints will withstand dy namic ef fects du-to transients, such as pump trips and salve trips, and that piping vibrations are within acceptab c les els.
- 1his includes ASNili instrumentation lines up to the first support in each of three orthogonal directions from the process pipe or equipment connection point, Revise Subsection 3.4.2,1.1 as follows:
Steady State Vibration S) stem vibrations resulting from flow disturbances fall into this category. Positive displacement pumps may cause such flow variations and vibrations, Since the exact nature of the flow disturbance is not know n prior to pump operation, no analysis is performed, if system sibration is evidenced during initial operation, the maximum amplitudes are measured and related to alternating stress intensity lesels based a the guidance of ANSI /ASME OM, (Reference 2) Part 3.
The APNK) preoperational sibration monitoring program includes appropriate safety-related instrument lines up to she-rwa-am4w the first support in each of three orthogonal directions from the procera pipe or equipment connection point.. The acceptance stan)wttm and contponent sibration is installed at the required locations.
(I irst plant only)
Reactor coolant sy stem hot functional testing is in progress.
ne inspection of the reactor s essel and internals will be performed following coohlow n f rom hot functional testing at full flow for greater than 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br />.
Test Method Verify the u uamic response of the reactor coolant sy stem. pressuriier surge line. passise core cooling lines, and pressunier relief and automatie depressurization piping and components dming both steady and transient flow conditions. (First plant only)
Mount transducers on the internals to acquire data during hot functional testing. (l'in,t plant only) l'ollowing cooldow n from hot f unctional t-sting, inspect the reactor sessel and internals and the document result.s. The areas included in the inspection are outlined in Subsection 3.9.2.4.
Performance Criteria The dy namic respo r. of the reactor coolant sy stem, pressuriirr surge line, passive core cooling lines, and pressuriter relief and automatie depressuritation piping and components is acceptable (First plant only)
The sibration data reduced and compared to allow able lesels. (f irst plant only) 1 l
An inspection of the reactor vessel and internals, follow mg 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> of full flow operation, is performed. and inspection results are acceptable.
210.58 2 W Westinghouse l
6 # W 7/- TOR ADDITIONAL INFORMATION i
Question 210.72 Section 3.4.7.1 of the SSAR states that the shroud assembly e,J the CRDM seismic support plate, w hich ar-both part of the integrated head package (IHP), are required to proside seismic restraint for the CRDM and the sabes and piping of the reactor head sent and are beh classified as AP6m equipment Class D, seismic Category I. It futher states that the shroud and seismic supmrt plate are categorized :.s intenening elements using the rules of the ASMI: CoJe, Section ill, Subsection NI', and are therefore not subject to the rules of Ni. In addition, Section 3.4.7.3 of the SSAR states that these two components are designed to the guidelines of AISC-N6uo. luga, ne staff does not agree with these classifications. Sections 3.4.4.3 and 5.4.12.1 of the SSAR state that the CRDM housing and the piping and equipment from the vessel head sent up to and including the second manoal isolation s ah e. respecth ely, are ASMii Clau 1.
In addition, the shroud is bolted to the ASMI: Clau 1 reactor pressure s essel head. Therefore, the statli position is that supports w hich proside seismie restraint for Clau i components and are also attached to Class I components cannot be categorized as intenening elements as defined in ASMF Subsection NF, but should be classified as AP600 Class C and constructed to all of the rules of Subsection NF.
I Revise Sections 3.4.7.1 and 3.4.7.3 and Sheet 38 of Table 3.2-3 to reflect this stati position.
Response
The SSAR will be resised to reflect the NRC staf f's position SSAR Revision:
Resise the first paragraph of Subsection 3.9.7.1 as follow s:
Components, including the shroud and control rod drive mechanism seismic support plate, required to pros ide seismic restraint for the control rod dris e mechanisms and the s ahes and piping of the reactor head vent are AP600 equipment Class CD, seismic Category I. The shroud and seismic support plate are designed in accordance wifb m4*yorunl4i 4nter+w inr+1emen4+4*4*y4h*4ulew4-the ASMi! Code, Section 111. Subsection NI requirements.
Themio,*r-lhe e+omtenen4+,-*r*-not--+,*ds*4-4*4ho *4"tuisement*
Resise the ficst paragraph of Subsecti n 3.4,7.3 as follow s:
The components of the integrated head package, w hich provide seismic support including the control roJ drise mechanism seismic support and the shroud, are designed using the ASMii Code. Section lil, Subsection NF.
puhlehim.-4--AlW-Nwa4 ux4-e
!!! unt-4he-A lW41*mmi-of44eet--Com 4m4 ion 4H+f#rwwe-4h liccause of the application of mechanistic pipe breal evaluations, the supporting elements do not : ve to be designed for loads due to a postulated break in a reactor coolant loop pipe. Pipes dow n to four-inch nominal diameter are esatuated using mechanistic pipe break criteria and the integrated head package is analyzed for movement of the reactor seuel due to a break of any pipe not qualifici for leak-before-break. See Subsection 3.6.3 for a discussion of the mechanistic pipe break requirements.
210.72-1 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Res ise Subsedion 3.4.M as Iollow s:
- 11. Deleted. 4pe.44keih* 4+. rah *-lWyncl+1 e**the m,*l&,* th*4144c.44aletwHel*4cd44m,4m*4+44u,4cae l#441.tiw4,neshwt4n4itute4444cel 44.n4n*4km4 AIM h-AlW-Nt.WM4X4r
- 12. Deleted. 41* meal-4444*c4 4'we.4m*4h.n/-14igleth44titionAn e+bwnle>4isus+4444cel44m4rm4h.n-Resise Table 3.2'3 (Sheet 3M of 107) as follow s:
Colnponent Al'600 Seistnic l'rincipal l)cAcrip h i d)('
( '111 %
Cole L' Ors ([ofist.Codt*
Conil110His RNS.A1Y Y60 11 CD l
ash 11!, N1' AIM'-4.uu CRI)N1 rool. LNG SilRotID RNSSil01 URI)\\1 SEISNilf SI'PI' l'I.ATl! II CD I
AShili. Nif AIM 4*)
5 210.72 2
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NRC REQUEST FOR ADDITIONAL INFORMATION I
Question 210.82 Resision i to WCAP-13054 lists an exception to Section I.al1)ol RG 3.10. that states that for electrical equipment.
the only dy nanne loads considered in testing are seismic loads, and that these seismit loads are not combine <! either by test or anal sis with other dy namic londo if there are any dynamie loads other than seismic that could af fect 3
either the cipiipment or the floor response for the equipment, these loads should be included in the equipment qualification either by test or analy sis. Resise this exception to provide a detailed basis for not including such load; Respont,o:
5% AR Subsec tion 3.10.2. requires that the efleet of dy namic loads in addition to seismic loads be censidered in the qualification of cl-ctrical equipment w here applicable. The S$AR addresses the NRC recommendation gis en in the Standard Resiew Plan, Section i.nll) to consider nonseismic dynamic loads as part of the equipment qualification process.
The position for Standard Resiew I'lan. Settion 3.10-1.all) will be revised in the nest resision of WCAP-13054 to remose the exception.
SSAR Revision N O N!!
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NRC REQUEST FOR ADDITIONAL INFORMATION n.
..,. -A Question 210.83 Revision I to WCAP-13054 lists an exception to Section I.a.(2) RG 3.10. that states that w hea performing seismic qualification of mecLanical and electrical equipmc.i by test, all accident loads are not superim[vned on the seismic loads. Revise this exception to dese% the types of accident loads that will not be i.uperimposed on the scismic loads.
Response
SSAR Subsection 3.10.2, requires that the effect of dynamie loads in addition to seismic loads be considered in the j
qualification of electrical equipment w here applicable. Section 3.10 2.2 states that actise mechanical equipment is 1
qualified by a combination of testing and analysis which -ddresses non seismie loads. if applicable. The SSAR addresses the NRC recommendation given in the Standard Review Plan, Section 1.a.(2) to consider nonseismic 1
(
dy namic loads as part of the equipment qualification process.
'the position for Standard Resiew Plan, Section 3.10-1.a.(2) will be revised in the next revision of WCAP-13054 to remos e the exception.
SSAR Revision: NON!!
W WestinEhouse
==
NRC REQUEST FOR ADDITIONAL. INFORMATION
-m Question 210.86 Resision I to WCAP-13054 lists Section 3 of Section 3.10 of the SRP as being not a part of the design process.
Sec' ion 3 states that complete and aud; table records of the seismic and dy namie qualification of equipment should be available and maintained by the applicant / licensee for the life of the plant. 'Ihe statf agrees that for the AP600, such documentation will not be available for design certification. How eser, the staf t's position is that the SSAR should state that the COL applicant should proside these records, which, in part, should consist of information similar to that in the sc.mple equipment qualification data package in Attachment A of Appendix 31) of the SSAR.
Revise the exception to Section 3.10 of the SRI'in WCAP-13054, and Section 3.10 of the SSAR to proside this statement.
Response
Section 3 of Section 3.10 of the SRP defines the documentation that will be necessary w hen safety related equipment (mechanical' electrical) qualification hs been cornpleted on ea-h piece of equipc.ient. Similar information is required by Section 8 of IEl!E 382-85 and Section 10 of lEliE 344-87 which serse as seismic qualiGemion criteria for the A P600. Section 3.10.4 of the SSAR states that AP600 documentation will satisfy reguhtory requirements.
Westinghouse will maintain the equipment qualification file to provide for review, audit, and inspection required for equipment supplied by Westionghouse. This file may serse as an example fo the Combined 1icense applicant.
'the Combined License applicant must s erify that the equipment quali6 cation file is maintained during the equipment selectien and procurement phase.
- 1he position on Section 3 of Section 3.10 cf the Standard Revie: Plan will be revised in the next revision of WC A P-13054 0
SSAR Revision:
'the first paragraph of 3.10.4 will be resised as follow s:
The results of teF., and analy ses serifying that the criteria established in Subsection 3.10.1 are satisfied, employing the qualitication methods described in Subsections 3.10.2 and 3.10.3, are included in the individual equipment qualification data packages and test reports. They are referenced in Table 3.11-1. The combined license applicant will s crify that the equipment qualification file is maintained during the equipnwnt selection and procurement phase.
Add a line to Table 1.8-1 as follow s:
210,86-1 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION j
b=
Table 1.8-1 (Sheet 3 of 8)
Summary of AP600 Plant interfaces With Remainder of Plant item Interface Interface Type Matching Section No.
Interface er Sub-Item sec: ion 3.13 Maintain the equipment qualincatio.1 file Requirement of Combined Licene 3.10.4 during the equipment selection and AlWK) applicant program procurement phase 210.86-2 Vj Westinghouse i
NRC REQUEST FOR ADDITIONAL INFORMATION
.=_y Question 210.92 Revision I to WCAP-13054 lists an cucption to Section I.ad6) of Section 3.10 of the SRP, that requires some clarification. It states that for an earthquale less than an SSli, each principle axis is simultaneousl) escited, and if no principal axis is evident, the equipment is positioned in the worst case orientation. Describe how the worst case orientation is determined if no principal axis is esident.
Response
The seismic testing methodology to be employed by the AP600 Project is in compliance with the testing reconunendation set forth in Standard Review Plan Section 3.10 subsection 1.aj6). This position is documented in Section li.5.1 of Appendis 3D to the SSAR. It is not feasible to define principal axes to qualify electrical equipment w hich houses electrical desices and components. %ese electric components may be mounted in the cabinet at various levels orientations and mounting details. These s ariabilities make it dilficult, if not impossible.
to choose the critical cabinet orientation. llecause of this it is industry practice to use a set of orientations in seismic testing. Section !!.5.1 of Appendix 3D to the SSAR defines how the NRC.ecommendations in SRP Section 3.10 subsection 1.a (6) are met.
The AP600 position for SRP Section 3.101.a.(6) will be revised to reflect this response in the next revision of WCA P-13054 SSAR Revision: N O N ii 210.92-1 l
W Westinghouse l
l
l NRC REQUEST FOR ADDITIONAL INFORMATION
.y_.-
Ouestion 210.93 Revision i to WCAP-13054 lists an exception to Section 1.c of Section 3.10 of the SRP, that states that the AlW!
desien implements II!Eli-323-1983 and III.li-344-1987. The last sentence states ths.t justification will be prosided if the test sequence is not specifically follow ed, e.g., aging by analysis. The statrs position is that the last sentence should be deleted, and exception" be replaced by " acceptable." If the last sentence is not deleted, describe how aging by analysis is accomplished.
Response
Age sensitise materials can be evaluated from a thermal and radiation aging standpoint if the aging characteristics of the materials are know n. Such an analysis can determine w hether significant degradations will occur under the espected Class 11! service conditions. If analysis concludes that insignificant degradation will occur and that the critical component will still perform its safety-related function, then thermal or radiation aging may not be necessary.
)
Similarly, performance of electrical or mech. cal operational cycling may not be required if existing design data demonstrates equipment durability greatly in excess of the estimated number of operating cycles for Class lli service.
In addition, past qualification tests, w hich included aging as part of the sequence, may provide sulficientjustification such that aging of similar equipment is not required. 'lhe exception to Section 1.c of Section 3.10 in the Standard Resiew Plan in WCAP-13054 will not be resised.
SSAR Revision: NON!!
I w wemanouse
NRC REQUEST FOR ADDITIONAL INFORMATION i
Question 210.94 Table 3.9-13 of the SSAR lists control rod drise mechanism (CRDSI) production tests and their respectise acceptance standard to ensure operational adequacy. In Section 3.9.4.4 of the SSAR. proside a discussion on how the functionality of the CRDM is ensured for seismic and LOCA loads.
Response
l Tests to s erify the control rod drit e mechanism trip times, w hen esposed to I.OCA loads, were performed and are documented in WCAP 8446 tReference 9 in SSAR Subsection 3.4.8). laboratory testing has been performed to demonstrate the ability of rod control cluster type drivelines to be inserted under various types of static and dynamic misalignments. The results obtained from these tests support the conclusion that operability can be provided for the AlWO during and subsequent to postulated seismic esents. A brief summary of the test is given in Reference 14 to be added to the SSAR as shown below.
Seismic testing has iwn performed to demonstrate the performance of the dnseline under conditions of dynamic excitation by a Westinghouse licensee. This testing w as performed using a single full-site driseline, including a control rod dris e mechanism and rod travel housing, guide tube. fuel assembly, drive rod and an rod control cluster style control rod. During this testing, both sinusoidal and seismic-like accelerations were applied to the driseline.
The testing was performed for a range of input acceleration levels and accelerometers were used to measure the response of the various components of the driveline. Rod drop time measurements were made before. during and after testing to confirm the operability of the control rods, input accelerations were applied at points corresponding to the fuel assembly sessel supports and the top of the rod trasel housing.1:or the sinusoidal excitation tests, the frequency of the input was varied to match the natural frequencies of the control rod drise mechanism and rod trasel housing. guide tehe and fuel assenibly. The results confirmed that the control rods could be inserted into the core.
l This conclusion was further seinforced by the tests w hich were performed using the seismic wase input w hich also l
demonstrated that the control rods could be inserted into the core under the seismic conditions.
I SSAR Revision:
Res ise the second paragraph of Subsection 3.9.4.4 as follow s:
To confirm the operational adequacy of the combination of fuel assembly, control rod drive mechanism, and rod cluster control assembly, functional test programs hase been conducted. These tests verify that the trip time l
achieved by the control rod drive mechanisms meets the design requirements -The+4*.w4ee F e;wied-in WGAP-844Hbferem+ Laboratory testing has been performed to demonstrate the ability of rod control cluster type drivelines to be inserted under various types of static and dynamic misalignments. The results obtained from these tests support the conclusion that operability can be assured for the AP600 during and subsequent to postulated seismic events. A summary of the seismic testing is given in Reference 14. Other tests have been reported in WCAP-8446 (Reference 9). The design of the control rod dris e mechanism osed in the AP600 is substantially the i
same as that tested.
Add a reference to Subsection 3.9.8 as follow s:
210.94-1 W Westinghouse t
=
NRC REQUEST FOR ADDITIONAL INFORMATION
- 14. Iliroshi Atiyanui and Makoto Watabe, "l' roving Test on the Seismie l<eliability of Nuclear l'ower l>lant l'WR Reactor Core Internals *, Transactions of The loth International Conference on Structural Mechanics in Reactor
'lechnology 14-18 August 1984. Anaheim. California, USA.
210.94-2 VJ Westinghouse 1
fJRC REQUEST FOR ADDITIOfdAL INFORMAT'OM Question 210.9G in Section 3.9.5 of the SSAR. discuss w hich part of the reactor internals (core support and other internals), and under w hat conditions, the criteria of Appendix I; of the ASN11! Section til Cale are applied. Since the Code does not ensure functionality, identify additional requirements used to ensure the safety f unction of internak.
Response
Section 3.9.5.3.1 and Section 3.9.5.3.2 include a discussion of the design basis requirements under w hich the criteria of Appendix l' are applied. The safety f unction of the internah, are auured by meeting allowable of upper barrel deflections and guide tube displacements. 'ihe reactor internak consist of core support structures. threadsd structural lasteners, and mternal structures are analy fed and compared to the allowable stresses of Appendix 17 Satisfying Appendix 1: requirements for core support structures and threaded structural fasteners and suceting the allow able deflection requirements in Table 3.9-14 and Set tion 3.9.5.3.2 pros ides the safety f unction of the internah.
Consequently, no additional requirements are necessary to ensure function of the internals.
SSAR Revision: N O N ii 210.96-1 W Wes1;nghouse
NRC REQUEST FOR A JDITIONAL INFORMATION 7
h, Ouestion i s ').99 the crow sectional drawings show n in I igures 3.4-5 and 3.4-6 of the SSAR lack detailed descriptions. It is unclear how dHierent parts of reactor internals are connected. In addition, identify the relatise locations of the internals to uch other and to the reactor s essel. Revise the figures to clarity the design detail shown.
Response-
'ihe attached l'igure 3.431 show s the reactor internals interface airangement identifying the position of each reactor internala component and the relatise position of the upper core support structure (l'igure 3.4-6) to lower reactor intern:Js (l'igure 3.4-5). 'Ihe lower reactor internals rests on the vessel ledge as show n in the figure. 'the upper en support structure (l'igure 3.4-6) rests at the same sewel ledge location on the top of hold dow n spring. 'ihe halddow n spring is between the core barrel flange and the upper support pla'e flange as show n in the figure. Iloth the assemblics are held together by the clamping force from the forty-five (45) 7" studs which clamp the upper head to the upper shell of the reactor s essel. 'the low er reactor internals is also supported h3 four support lugs w elded to the bottom head of the vessel. 'lhe low er core support plate of low er reactor internals rests on these lugs.
SSAR Revision:
Add a new subsection and l'igure 3 4 8 for Reartor Internals interface as shown in the following:
3.9 1 1,4 Reactor Internals Interface Arrangement:
l'igure 3.9-8 shows the arrangement of reactor internals components shown in l'igures 3.45 & 3.96 and their
(
relative position in the reactor vessel. As shown in the figure, the lower reactor internal (figure 3.9 5) rests on the l
vessel ledge. 'Ihe upper core support structure (ligure 3.9-6) also rests at the same location, but on the top of a large compression spring (hold down spring). The hold down spring is between the upper support plate flange and the core barrel flange as shown in the figure, lioth the assemblies are held together by reactor vessel closure studs, which clamp the upper head to upper shell of reactor vessel. The lower reactor internal is also supported by four support lugs welded to the bottom head of reactor vessel. 'Ihe lower core support plate rests on these lugs.
210.09-1 W Westinghouse 1
l 1
l
NRC REQUEST FOR ADDITIONAL INFORMATION
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NRC REQUEST FOR ADDITIONAL INFORMATION y.._._,
Question 210.101 Supplement the response to Q210.17 dated January H.1993 by providing key dimensions of the reactor veuel and supports and by specifying major design parameters (e.g., temperatures, pressure, etc.) of the react <,r vessel and internals.
Response
1he following is the design information retiuested on Reactor Vessel and intemals.
I eactor Vessel:
Design Pressure ( psi ):
See Table 5 3-5 Design Temperature ( 'l ):
See Table 5.3-5 Rewtor Vewel Key Disnensions:
(1 igure 5.3.4-l attac hed)
Vessel I.D 157.(X) in.
Vessel Support Distance i 14.(X) in.
f rot.: Vessel Center 1ine inlet Noule I.D 22.(K) in.
Outlet Nonle 1.D 31.(X) in.
Reactor Internals:
See 3SAR Subsection 3.9.5.2.5 below SSAR Revision:
Add section 3.9.5.2.5 as follow s:
3.9.5.2.5 Design Conditions Design Pressure Differential 35 psig Design Temperature 650 "F Add 1:igure 5.3.4-1 as attached.
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210,101-1 W Westinghouse l
NRC REOUEST FOR ADDITIONAL INFORMATION 1
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)
NRC REQUEST FOR ADDITIONAL. INFORMATION Ouestion 210.106 The AP6(W1 plant life design objectis e is 60 y ears. This propos;d design life raises questions relatis e to the margins asailable in the current ASME Fatigue Design runes. These margins were established abnost 30 years ago and were obtained from best fit cunes of fatigue test data by applying a factor of either two on stress or twenty on cycles, w hithes er was more conservative at each point. These factors w cre originally intended to cos er such effects as ensironment, site effect, and scatter of data. Ilowes er, based on limited data currently available, the staf f belieses that these margins may not be suf ficient to account for variations in the original fatigue test data due to various environmental effects. The ASN1ti Code cunes may not be resised for many y can.1herefore, the statf s position is that until these cunes are revised, all Al.WR's should include a proposed approach for accounting for environmental effects in the fatigue analpes for all ASME Class I sptems, components, and equipment and for the designs of ah class 2, and 3 sptems, coruponents, and equipment that are subjected to ey clic loadings,induding operating sibration loads and thermal transient effects of a magnitude and'or duration so severe that the 60-year life car.not he arated by the required Code analpes. In Section 3.4.3 of the SSAR, prov'.le a proposed approach to resohing ths concern for the AP6(M).
Response
The margins available in the current ASMl! Fatigue Design Curves are conservatis e.
The margins are intended to cour elfects of ensironment, size, and scatter of data, are two on stress or 20 on cycles. Industry has n,cd these raargins for nuclear power plant application for the last 30 years and found no esidence of f atigue faihire attributed to inadequate ASME design curves. Our understanding is that the source of the staff concern is mainly from Reference 210.106 l. With respcet to Reference 210.106-1 the following comments can be made: (I)1he oxygen content (8(K) ppb) data used by Reference 210.106-1 has limited, if any, applicability to the AP600 with MK) to 200 ppb osygen. (2) Reference 210.106-1 mises alllow strain rate and high strain rete data. Essentiahy, it applies the same curve for all esents, including the rapid rise time events w here there are no environmental effects. This is an oserly consenative approach that unrealistically applies the environmental penalty on all events. (3) The Pressure Vessel Research Council (PYRC) progress report by Sumio Yukaw a (Reference 210.106 2) concluded that the only case for consideration of ensironmental ef fects on fatigue design curves may be carbon steel and low alloy steel in certain llWR ensironments. No concern was expressed regarding stainless steel in PWR or llWR ensironments, or carbon steel or low alloy steel in PWR environments.
ilased on above discussion, AP600 components are designed according to the current ASME Code Section 111, Fatigue Design Curves (1984 Edition through 1939 Addenda) and no change in the SSAR is required.
References l
l 210.106 1 NUREG-5944, Interim Fatigue Design Cun es for carbon, l.ow Alloy, and Austenitic Stainless Steels in LWR Environments. April,1993.
210.106-2 ASME PVRC prcgress report to lloud of National Codes end Standards (IINCS) " Evaluation of
(
Fatigue Cunes in Sretions HI and XI in light of current world wide data", by Sumio Yukawa, on l
NRC REQUEST FOR ADDITIONAL INFORMATION behalf of the h t7 Ste ring Committee on Cyclic 1.ife and linvironmental fillects in Nuclear Applic ations", date 4'.4 SSAR Revision: NON!!
I l
i 210.106-2 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Question 220.61 It is stated in Section 3.8.2.1.1 of the SSAR (pg. 3.M-1) that a flexible watertight and airtight seal is pnnided at filevation 123' 3") betw een the containnent shell and the shield building. I'rovide clarification for the follow ing concerns:
What is the sa ety class and scismic category of this scal?
f a.
h.
if the seal is a wici) class item, explain how the seal will perform its f unction w ben the containment shell is displaced laterally inward and outward at the base of the foundation.
Response: (Revision 1)
Note: the correct elevation of the seal is 132' 3" as stated in Section 3.8.2.1.1 of the SSAR. 'the flexible w ater tight seal is utilized to seal against water leakage from the upper annulus into the middle annuhis. 'lhe seal also serses to maintain an intact holdup volume s ithin the mitidle annulus for containment leakage of contaminants follow ing a sesere accideut scenario
'lhe seal is designated as nonsafety-related and non seismic; it is not relied upon to mitigate any design a.
basis es ents.
h.
'lhe seal is, howeser, Jesigned to acconunodate esents resulting in containment temperature :md pressure excursions which result in lateral shell movement, inward or outw ard, as a result of normal operation.
design ba is accidents and sesere accident scenarios. 'lhe esents considered for the seal design are those discussed in SSAR Chapter 15 and the ses cre accident scenarios are those that do not exceed ASM E service l
level C limits for the containment, 1
SSAR Revision: N O Nii 220.61(R1)-1 W Westinghouse l
NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 220.91 After resiewing Westinghouse's March 24. 1994 response to Q220.32, the staf f has determined that it needs additional information pettaining to the fe' lowing issues:
Calculation of critica[preuure (CI') of 161 psjgjor the 16-ft diametfreguipment hatch cover.
i a.
A critical pressure of 161 psig im the 16 it-d.ameter equipment hatch cover was calculated by Westinghouse using the formula in Code Case N-284. T'as formula is based on a cylindrical shell. Ily using the formula f rom the Structural Analnint Shelt ny llaker et al. (pp. 253 254, AlcGraw liill,1972L de CP is found to be 152.8 psig. 'Ihe stalf beli, tes this formula is mor appropriate for this application. since it is based only on the spherical cap cmers (sersus a cylindrical shell), which proside s additional conservatism. Proside justification for the use of Code Case N-2H4 to determine the critical pressure, b.
Appljentiop of Code Cas.r_N;H4 to the instabiljtnanalysis of the 16 f t-diameter unstif fened equiranent hatch W)A The 16 ft-diameter unstif fened equipment hatch cover is subjected to esternal pressure. N284 is applicable to 8 l } local buckling of the shell plate betw een stiffening elements. (2) buckling betu een circumferential stif f eners of combined shell plate and attached meridional stiffeners, and (3) general instability or userall collapse of the combined shell and stif fening sy stem. On the basis of the staff's interpretation of N284, the unstiffened shell for the general instability should be cosered under Nii-3222, unless justified otherwise. The attached *.rble clarifies the stall position on the applicability of Nii-3222 and Code Case N284. Proside justification for not using NI.-3222.
1 in Section 10.2.5 of the APo00 PRA, the peak containment pressure is predicted to be 95 psia (80.3 psigt i
SI:CY 93-087, " Policy. Technical, and 1.ieensing issues Pertaining to livolutionary and Ads anced Light Water Reacter Designs." states that the containment stresses should r.ot esceed ASMl! Service 1.csel C limits for approsin ately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the onset of core damage under the severs accident challenge. liased on the abose discuwion the inel C Sersice 1.imit should be 61.1 psig (152.8 psig/2.5) for the 16-it-diameter equipment hatch cover instead of the 96 psig (161 psig/l.6:, salue indicated in the March 24, 1994 response to 'his RAl. Address this issue.
l l
220.91-1 W Westinghouse l
l l
l l
NRC REQUEST FOR ADDITIONAL INFORMATION
,.- -.-A Steel Containment Acceptance Cnteria for Local and Global Bucklings Global Buckling l
Local Buckhng NE 3100, -3200 N 284 N 284 i
- 1. Unstif fened Shells
- 1. Stif fened Shells
- 1. Unstif fened Shells (Nii-31(x))
- a. lixternal Pressure
- 2. Stif fened Shells
- a. lixternal Pressure
- b. Asial Compression
- a. Internal Pressure
- b. Axial Compression
- 2. l' actor of Safety" (Nii-3200)
- Symmetric
- Symmetric
- a. les el A: 3.0
- Unsymmetric
- Unsy mmetric
- b. level 11: 3.0
- d. Reduction Factors
(. Reduction l' actors
- c. lx tel C: 2.5
- Capacity
- Capacity
- d. lesel D: 2.0
- Plasticity
- Plasticity
- e. I: actor of Safety
- d. l' actor of Safety
- level A: 2.4
- lxsel A: 2.0
- I n e! B: 2.4
- tes el H: 2.0
- 1.xvel C: 2.0
- level C: 1.67
- level D: 1.6
- level D: 1.34
- Definitions:
Capacity reduction factor:
Accounts for the effects of imperfections and nonlinearity in geometry and boundary conditions.
Plasticity reduction tactor:
Accounts for the nonlinearity in material properties.
To be applied af ter effects w hich can influence buckling are considered.
See Nil 3222. I for detail.
220.91-2 W Westil;ghouse
1 i
NRC REQUEST FOR ADDITIONAL. INFORMATION
. =. =
Retponse:
Ihe dilference in critical pressures calculated by Westinghouse and by NRC staff is due to the dif ferent capacity a.
reduction f actors, designated by a and f( A), that w ere applied to the classical buckling pressures for a complete sphere. Ihe formula for the classical buckling pressure used by Westinghouse and the NRC staff are identical.
1he critical pressure calculated by Westinghouse is based on the ASN11i Code Case N-284, whereas the critical pressure calculated by NRC staf f is based on the formula given in *the Structural Analysis of Shells" by liaker et al. tpp 253-254, McGraw-liill,1972) In the AShill Code Case, the capacity reduction factor is a function of the unsupported length factor (M). whereas in the reference used by the NRC staf f it is a function of a geometry parameter ( A).
1he formulas for the theoretical (classical) buckling capacity of spherical shells subjected to external pressure and of c)linders subjected to axial compression are identical t f rom
- Theory of lilastic Stability", by Timoshenko et al.). The theoretical capacity is then reduced by the capacity reduction factor to obtain the critical buckling pressure. The capacity reduction factor (a) obtained from ASMi! Code Case N 284 and used by Westinghouse is based on spherical shell test data. Code Case N-284 was deseloped by the code committee based on detailed review and evaluation of test data. This test data includes more data than w as asailable in 1972 w hen llaker et al. prepared their reconunendations. The ratio of test buckling stress to classical (theoretical) buckling stress for spherical shells / caps uas plotted as a function of the non-dimensional unsupported length along the shell (M = ldRt, where L is the unsupported length along the spherical shell, R is the radius of the shell, and I is the thickness of the shell). 1hese plots are shown in the response to R AI 220.32. The lower bound cune based on this test data is used in Code Case N-284 (see I*igure 1512-1 of Code Case N-284). 1he capacity reduction f actor (n) was found from this con e and then applied to the classical buckling pressure to determine the critical buckling pressure. As this approach was based on the low er bound curve of N 284 test data, and this code case is based on a more recent res iew of available data than w as available to Raker et al. in 1972, the critical buckling pressure of 161 psig calculated by Westinghouse is considered appropriate.
b.
The 16-ft diameter unstiffened equipment hatch cos er subjected to external pressure falls in the category oflocal buckling of the shell plate between stiffening elements The Hange of the cover acts as a stiffening element around the periphery af the spherical cap. Therefore, the use of ASME Code Case N-284 with its factors of safety for hical buckling is appropriate.1he Ixvel C Service Limit for the 16-ft-diameter equipment hatch cover is 96 psig (161 psig/1.67) as given in the SSAR and the design satisfies position IJ. on containment performance in SECY-93487, Pon y, Technical, and Licensing issues Pertaining to Evolutionary and Advanced Light Water Reactor Designs" I
SSAR Revision: NONE l
1 l
i l
l 220.91-3 l
W Westinghouse i
l l
NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Ouestion 230.11 Section 3.7.2 of the SSAR presides a scry general design requittment for Category 11 structures by stating that
" Seismic Category II building structures are designed and'or physically arranged so that the safe shutdown carthquake (SSin could not cause unacceptable structural interaction with or failure of their adjacent seismic i
Category I struc tures, o stems, and components." Proside detailed analy sis methods and design criteria that will be used to inert this general design requirement. I:or cumple, w hat seistnie an-J) sis will be performed for Category ll structures?
Response: (Revision 1)
See response to RAI 230.54 The SSAR is resised so that the de4pn of seismie Category 11 structures uses the same input and acceptance criteria as those for the seismic Category I structures.
SSAR Revision:
See the SSAR revisions for the response to RAI 230 54 l
1 l
l 230.11(R1)-1 3 W85" gh0USB i
NRC REQUEST FOR ADDITIONAL INFORMATION lU "i. n i,-
g Question 230 54 Acttetling to the ileillution in Sectioli.4.7 of the SSAl{, the scisinie.tnaly sis athl tiesiyn of the seisntic C;ttepoly li strih tures ;hljacclit to the nthleetr islathl shouhl be perlotinetl usilly tile s.tlite liiput arial aet eplance criteri;t as those foi the scisinis Category I siructures. Justily the use of the Zone 4 requirernents 01 il.e l'nitorin fluiliting Umle 11ill(7 hir the ;in;tl) sis anil tiesif fs of these strth itsres.
Response
The SSAlt will bc resisni;is sliown below so tliat the tiesign of sessritic Category 11 structures itses the saisie inpiit ant! act eptance ciilesia as those for the seisniic Categor) I stith tures 'Ihis lesision also ilestribes the tiesien of the finn-seisniic structures.til).icerit to the riuclear islan41.
SSAR Revisions: See nest pages f
W WestinEhouse
l I
NRC REQUEST FOR ADDITIONAL INFORMATION 1
1.2.5 Annex l and 11 Buildings l
l lluilding l' unction I
l The annex ! buikhng (Figures 1.2-23 through 1.2-26) provides the main personnel entrance to the power generation complex. It includes accessways for personnel and equipment to the clean areas of the nuclear island in j
ihe auxiliary building and to the radiological control area via tb-annex 11 building. The buihling includes the health i
physics lacilities for the control of entry to and exit from the radiological cantrol area as well as personnel support facilities such as lunch and locker nxuns. The building also contains the non-lE ac and de electrie power systems, other electrical espiipment, the technical support center and various heat, ventilation and air conditioning systems.
j No safety-related etpiipment is located in the annex 1 or annes 11 buildines.
The annex il buihhng IFigures 1.2-23 through I.2-26) includes the health physics f acilities and provides personnel and equipment acetssways to and f rom the contaimnem buihling and the rest of the radiological control area via the auxiliary buikhng. Provided are large, direct accessways to the upper and lower equipment hatches of the containment buihling for personnel access during outages, for laree equipment entry and exit and for new tuel I
transport directly to the operating det L3 of the containment or the tuel handling buil hug. The building also includes a hot machine shop servicing radiological control area equipment. The hot mact.ine shop includes decontamination f acilities incluthng a portable decont.auination system that may be used for decontamination operations throughout the nuclear island.
Cin il St ruct ural l'eatures The annex 1 and h buildings are non-seismie stn.ctures designed to the Uniform Buihling Code. They are designed so they will nat tail in a manner w hich would damage safety-related structures. No prmection against missile penetration is required, llowever, certain areas of the buihiing. such as the hot machine shop and the technical support center, are provided with shickling for protection against low level radiation fro a either internal decontamination systems or inun external sources under accident conditions. This is accomplished by either reinforced concrete walls or reinforced masonry walls.
The annex ! and 11 huikhngs are steel framed structures with insulated metal siding. Fhior and not slabs are remforced concrete supported by metal decling. These act as diaphragms to transmit horizontal loads to sidewall bracing. Indivi.keal+einbuteduwsetemalqwvhk4e-hmndalientimwluS41mimiklinpCombined spread j
hiotings provide the foundation for the annex buildings.
l l
1.2.6 Diese! Generator Building l
fluilding l' unction The diesel rencrator building (Figures 1.2-27 and 1.2-28) houses two identical slide along diesel generators l
s, narated by a three hour tire wall. These generators provide backup luwer f or plant operation in case of disruption of nonnal power sources. No safety-related equipment is hicated in the diesel generator building.
Cin ilStructural l'eatures l
l i
230.54-2 W WestinEhouse l
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NRC REQUEST FOR ADDITIONAL INFORMATION 1
The diesel generator buildmy houses the two diesel generators and their associated he.iting. vent.hition and air contlitioning equipment. none of which are required for the sale shuttlow n of the plant. Accordmgly. the buildmg is designed as a non+eismicsafety-relateil structure subject to wind and seismicenvironmental huds in accordance with the Unifortn fluilding Code.
Thei hng is a single story steel inuned structure with insulated metal siding. The root is composed of a metal deck supporting a concrete slab and serves as a hori/ontal diaphragm to transmit latend loads to sidewall bracing and thereby to the foundation.
The foundation consists of a reinforced concrete mat. The diesel generators are skid mounted and rest on vibration isolators supported directly f rom the mal. spread i+witiny+with ewuns-linytnelelwstm vanwmd thoperimeter
.+the-boilding.-A o unaetr+1ah 49. youle wt ve. as the lir t 4h orApoesi+wif + int +-ure-prow ledletween Iliediesel yenerat >r4+nitidati4Hi+-4esi48-4heslaiv4w+-yntili'anil-lwatwien-ilieslah +tnil-ilis'yradelwim+.
1 1.2.7 Radwaste Building l
lluilding Function The radwaste buihling include 3 lacihties for the following f unctions. No saf ety-related equipment is hicated in the radwaste building.
Temporary storage of various categories of plant wastes Contaminated laundry cleaning, monitormg. repairing and storage Respirator cleaning, testing. repairing and storage Dry waste processing and storage Clean dry waste veritication and removal trom radiological control area Ila/ardous/mited waste sampling and accumulation Cartndge liNr packaging and storage Spent resin and tiller bed media storage / packaging and packaged resin storage Detergent waste storage and pmcessing Chemical waste storage and treatment by mobile systems Empty waste container receiving and storage l
Storage and loading packaged wastes for shipment 230.54-3 l
W Westinghouse i
NRC REQUEST FOR ADDITIONAL INFORMATION Cis il Ntructural I?catures The radwaste building general arrangement is show n on Figure 1.2-29. The radwaste building is two septrate structures. The portion east of Une GR is a non-seismic structure designed in accordance with the Unitonn Building Code. The porhon west of line GR is a high bay area for storage and handling of. spent resin and !iquid waste. This portion is designed as a seismic Category 11 structure. The spent resin storace and processing area and liquid radw a3te storage areas are designed to contain any liquid or resin spills resulting f rom a sale shutdow n earthquake.
The loundation for the seismic Category Il structure is a reinforced concrete mal which is separated from the combined spread hiotings of the non-seismic structure 1.2.8 Turbine Building l
l lluilding 12 unction The imbine buikhng houses the main turbine generator and associated fluid and electrical systems. 11 provides weather protection for the laydow n and maintenance of major turbme/genciator components, The turbine buihhng also houses the makeup water purification system. No salety-related equipment is hicated in the turbine building.
Cisil' Structure Features The turbine buihhng, d. 'wn on Figures 1.200 through 1.244, is a steel column and beam structure. The turbine build.pq pround floor (structural mal) is a reinforced concrete stab. The turbine building is a non-seismic structure designed for wind and seismie loads in acconiance with the Umform Buihling Code.
The turbine-generator is low-tuned by means of spring supports. The design consists of a reinf orced concrete deck mounted on sprmps. The springs are supported m a structural steel framework that forms an integral part of the turbine building structural system. L.ateral bracing serves to provide lateral support for the buihimp as weP as the turbine-generator support. The spring-supported concept isolates dynamically the turbine-generator deck f rom the remainder of the struClure for operating IrequenCies, thtis allowing for an integrated structure below the deck.
t his istudes an integrated reinforced concrete foundation mat that supports both the turbine generator and the buihling. The condenser is attached rigidly to the low pressure turbine exhaust and is supported on springs. The loumLition f or the entire buildmg is a reinforced concrete mat. -The4mihling-iw4eismieCatemyil-imihling.
230.54-4 W WestinEhouse
NRC REQUEST FOR ADDITIONAL INFORMATION if iiji i
a; Table 3.2-2 Seismic Classification of Building Structures Structure Category Nuclear Islarul C-1 liasemat Containment Interior Shiehl lluilding Auxiliary fluihling Containment Air llalile Containment Vemel C-I Plant vent and stair structure C-Il Turbine fluilding NSC-Il Annex iluilt ing i NSC-41 i
Annex lluilding 11 NSC-il Radwaste fluilding
- Columns AR - GR NS
- Columns GR - IR C-Il Diesel-Generator Eluilding NS Circulating Water Ptunphouse and Towers NS C Seismic Category i C-Il-Seismic Category 11 NS-Non.scismie Note:
Within the broad delinition of seismic Category I and 11 structures, these buildings contain members and structural subsystems the lailure of which would not impair the capability for safe shutdown. Examples of such systems would be elevators, stairwells not required for access in the event of a postulateil earthquake, and nonstructural partitions in nonsdcty-related areas. These substructures are classitied as non-seismic, W WestinEhouse
NRC REQUEST FOR ADDITIONAL INFORMATION l
3.7.2.8 Interaction of Seismic Category 11 and Non-seismic Structures with Seismic Catepon. Sirectures. Systems or Com;wments Notescismic structures are evaluated to detennine that their seismie resgwmse shies not lirechale the safety lunctions of seismic Category I structureser systems or ctunponents. This is accomphshed by satislying one of the following:
The collapse of the non-seismic structme will not cause the non-seisnue structure to strike a seismic Category I stnkture system or component.
'T he colhpse of the non-seismic structure will not impair the integrity of wismic Category I structures, systems or components The structure is chtssified as scismic Category 11 and i< :culy/ed and designed to pres ent i,s collapse under the safe shuidown earthquake. The structure is analyst %r the safe shutdown earthquake using the same meth(wls as are used for seismic Category i stroetures. Lt.J tactors and load combinations for concrete structures are in acconiance with ACI 318, except that the load factor for the SSE is taken as 1.0.
Allowable stresses in steel structures are in accordance with AlSC with a 600 increase pennitted for the SSE instead of the one third increase.
Tla%4nw4me+that art %ulikiemlys hise to the mwlear island ilwit sheir-s:olhipss% ouhinited+afetylmwtions areabeamwel-malli-buildiny+cahe-hiplebaymeaa441w-rwiwaqehuikhngaeohemn line+GWt+il4*+showwin R y me'l :2-2Wrend -the4orbinebuihling JI'hese st nwure+are44assi fiale+seismieCat tv erydl,tnil areanalp ed anel ele igned--to prevent 4heiFlaihrre-unilerala'*afs%hutdown-eanlatuakt'SeisodeCeittigerv-41+1rus turs*4+++.lswigswil
+acoordans e-willFACI J IK4oFomerete-stow sme+and-willFAISC4oFMeektow4meeSeismie load +areitnalpol i
aceor linp+ r the41nifomFHuihliny-G wle-4 Ref eren+2 h4+ me4rfatuir*9nent *-usingiusiutportmw o bw tor 44 -I A TheJ melseismiev+me4rwtor-i-thM-whieh-4+omsi4enFwillethe4Mgsera-twgi+%I-aeaderati m-ofale+afe simitloWu earthtpiaker SeismieCateyory41-unetort*-aro physioilly i+rrangssl +.vihat-sliesals%lmtilown ssirtlht*H*kt 4SSM-*l+ ** ""i tausellem4oumt.wbad went*ismieCweyoryl+tmeture*-Tir'minimunFspace-reituireillietweim+tnxami+4o t
avoidsamladi+obtailwtl by perlonningeitbeFa time 40storyof 4t resp.m e+[witnunitnalysi4for4Nwh+4nNture'.-The disidat ement + resultint i n mbthese,u miyw+ areaihled,4wolutely,it.espd vakmt elevation +imtwoedpwenkt ruam es in-or.lerao41etermine ilw-minimum +pacee stuited Iw9wn9641*e tw*++1nwime--The-mialyw++44te-mwleardshawl an't wrlonned -using-t he+eismie-i ntmba+ defined da-SulwaaionMi-Th 'analy.e*4441*'seismie-C,ttey+ wv-il sinwtua~are-perlonned-nsing-41*useismie-inpubawlethwd-in4he4:nitorneRuihling Co.ks l
The structures adacent to the nuclear island are the annex 1 and 11 buildings, the radwa.ste building, and l
the turbine building. The high bay area of the radwaste building (column lines GR to IR, as shown in Hgure 1,2-2W is classified as seismie Category II. The remaining portion of the radwaste building is a single story steel frame building. This portion is non seismic and is designed to the seismic requirements of the Unifonn Bulding Cmle. If l
it were to collapse in the safe shutdowr. carthquake, it would not impair the integrity of the reinforced concrete nuclear island not the high bay area of the radwaste building, i
230.54-6 W westinohouse o
l l
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NRC REQUEST FOR ADDITIONAL INFORMATION
...@H.hi The annex buildmg and the tsrbine building are classified as non-seismic and are analy/ed to demonstrate that their collapse under the safe shutdown earthquake will not impair the integrity of seismic Category I structures or comi.onents. They are arranged to provide suf ficient clearance from the nuclear island so that there is no impact between major structural elements of the buildmps. The annex bmlding and turbine building are concentrically braced steel f rame structures with the bracing resisting seismic loads in the direction towards the nuclear island stronger than that in the direction away f rom the nuclear island. This increased strength in the direction towards the nuclear island is such that any collapse due to seismic loads is not in the direction of the nuclear island. The design of the buildmps meet the following criteria:
The struttural steel framing is designed so that the stresses associated with the dead and live load are within o
code allowable stresses (S). The stresses resuhing from the 110 mph wind are maintained to stress levels below 1.33 S. The seismie loads delined by the Uniform Buildmg Code (Reference 2) meet the stress levels delined as 1.33 S using seismie input for Zone 2A with an hnportance Factor of 1.25.
The lateral load resisting system resisting translation towards the nuclear island is designed to have a o
strength 503 greater than the lateralload resisting system resisting translation away from the nuclear island.
The minimum clearance between the major structural elements of the non-seismic building and the seismic o
Category I buihlings is 1.5 times the absolute sum of the SSE displacements of the two buildings. The dellection of the nuclear island is calcalated by clastic analysis. The denection of the non-seismie building is calculaled in a simplilied nonlinear analysis using the Ap600 SSE input time history given in Subsection 3.7.1.
Steel structural bracing connections are designed with sutlicient strength to develop tensile yield in the o
bracing before the connection fails. This assures that the structure has sufficient ductitity to satisly the I
assumptions made in the inelastic analyses.
l o
The non-seismic buildings are designed for the tornado (wind speed = 300 mph) such that the stresses in the bracing resisting loads in the direction of the nuclear island are less than 1.7 times the AISC allowables.
S; ding is pennitted to blow oft during the tonnlo.
Interfacing structures, systems and components crossing the boundary between the non-seismic be lding and i
o the seismic Category ' nuclear ishind are designed such that their failure due to large relative displacements of the buildings do not jeopardire the required function of safety related items. The walls of the nuelcar island are 24 inch thick reinforced concrete walls. The maximum loads that these interfacing items can I
impose on the nuclear ishmd, corresponding to crushing or other failure of the item, are evaluated to l
demonstrate that they are smaller than loads, such as tonudo missile impact, for which the nuclear island i
is designed.
I I
230.54-7 W Westingh0'JSe l
1
NRC REQUEST FOR ADDITIONAL INFORMATION
!!f tim 3,7.2 Seismic System Analysis Itevise third uragraph of Section 3.7.2 as f ollow s:
t Seismic Category I bmhhng structures are on the nuclear island. Other buikling structures are classtlied non-seismic or seismic Category 11. Non-seismic structures are anal ><ed and tiesigned for seismic loads acconting to the (!nif orm Iluihling Cale (iteterence 2) requiicments f or Zone 2A. Seismic Category 11 buihhng structures are desiyawsi-aiallor-phy ~icallyainanyed++--Ilul4he+ ate +hutibiwn warthspuke4SSI4o>uki-not+au e-mucwptaNe si nn t ural-interact ion -w it h -oeul ailm e-of - t heir-a. I biwnt -seisenie Category 4+t rik tiirs%--systeinscaewi +oentionentsc--
analysed for the safe shutdown earthquake using the same methals as are used for seismic Category I structures.
Load lactors and loail ownhinations for seismic Category !! concrete structures are in accordance with ACI 318, except that the load factor for the SSE is taken as In Allowable stresses in seismic Category 11 steel structures are in accontance with AISC with a bO'M increase penuitted for the SSE insteail of the one third increase.
230.54-8 W Westing, house 1
l
NRC REQUEST FOR ADDITIONAL INFORMATION 1
~
U Response Revision 1 Ouestion 230.58 Provide the following information pertaining to the high frequency mmles of the structures:
Prmide justification to demonstrate that the time steps used in the time-history 3eismic analyses are small a.
enough to account for the high-frequency modes that have significant mass participation factors.
b.
Make " missing mass" corrections to the seismic analyses (horizontal as well as vertical) w here signilkant high-frequency modes were left out. Note that the seismic forces computed without such " missing mass" correction (if applicable) would result in underprediction (example: a foundation mat design w here seismic forces were used in the equivalent static analy sis).
Response: (Revision 1)
Seismic analyses for the hard rock soil case are performed using mcde superposition time history analy sis w hich might be affected by the effect of high frequency modes of response. Seismic analy sis for the soft rock and the soft-to-medium soil cases are perfomed using the complet frequency response analysis method which considers all masses of the model. and therefore additional consideration is not required.
Ibe-4he-hanWhdW+.+,-**.d++iywtwwitien4ime-tJ 4ery-mmly e -er+-twfornwd-4o-*44*h+-evelera4kuw dk.t aeement-mwWmidw4m+-re twwwe --fee-the-mekw-44and-+tmetum--**eep44he-memtw-kmv -fw-41w d
eensaimmit4ntena*Wi net um-494heem*l*-*uf wtwwitkc4ime4d*4*y-analy+.e -edi-unleWd+4m tuem'y-up-to 444 mue-im4mleir--The-eu nmladve-mada !
,. :a.klered4nah+eemlywe.-ere-M-Aryawt4mga}m nw'-
s!nad-nodal-maele++gnale4ocaheto+4h-wutlw4he.
. ' and4h++ertieal4im4 ion --+c.peetively, 14enee-dee4dgh4mpwney-n*wle +f4he+4wour e, "a nt-4h*w.*4 aim an44ntemaktetermem>4*leml4o4w.
in4 ped 4ienate--Thetime 'ep =f M4%eeoml-u ed4n4he-amdyn
- !cqm*4+amt4.-gamally-neeeptwidoc-em*le wigwpo44kmaime4d*4*r.w+mdy e*-up4* u-Rh The synthetic time histories were based on time steps of 0.01 seconds. As shown in SSAR Figun s 3.7.1-6 to 3.7. l-
- 8. the spectra for these time hi toriu match the design spectra and satisfy the requirement in Standard Review Plan 3.7.1. For time history analysis of structures having significant modes up to 33 Hertz, the input time histories are I
interpolated to give time steps of 0.005 seconds.
The containment internal structures is a relatisely stiff structure with significant response of high frequency mmles i
as show n in Table 3.7.2-3 of the SSAR. Mode superposition time history analysis for the containment inwmal structures includes high-frequency modes to bring the cumulative participated mass up to an acceptable level, and the time step size of 0.005 second becomes unacceptable. Tnerefore, as presented in the third paragraph of Subsection 3.7.2.2 of the SSAR, the member forces for the 3-dimensional lumped-mass stick model of the containment internal structures for the hard rock soil case are calculated by the response spectrum anal) sis including l
the high frequency responses using the procedure given in Appendix A to SRP 3.7.2, Revision 2.
Hence the
" missing mass" correction is considered in the seismic forces computed from the containment internal structures.
==
l NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 l'or the basemat design, seismic forces and accelera; ions for the soft rock and toft-to-medium soil properties are obtained from the SASSI analyses of the 3D lumped mass stick models. In SASSI analysis
- missing mass
- is not a concern and corrections are not required. Seistnic forces and accelerations for the hard roci soil pratile are obtained as described in the previous paragraphs and IN " missing mass" cor.ection has been considered.
SSAR Revision: NONii 230.58(R1 F2 W Westinghouse
NRC REQlJEST FOR ADDITIONAL INFORMATION
!!!O M Response Revision 1
~
W Question 230.63 The soil column pmpenies for horizontal and venical (P-wave) models are not consistent. Specifically, (a) the damping ratio lor S A P wave motions are ditterent, and (b. Poisson's ratio for soils above the ground water table appear to be too high. Provide an explanation to justif y ta) the use of same properties for the hori/ontal and vertient nunlels, and (b) the use of a high Poisson's ratio in the analysis.
Response
The strain-compatible soil / rock shear nunlulus and damping Oor shear wase) were obtained trom the averaec properties of two Sil AKE analysis results of the respective soil! rock profile using 111 and 112 time histories as described in Section 2A.4 of the SSAR. These properties shown in Tables 2A-9 through 2 A-12 were subsequently used in the SSI analysis. Along with these properties, the P-wave selocity and the damping associated for P-wave were obtained and used m the SSI analyses as described below, a.
The generie strain-dependent shear wase damping curves for soillmck matenals were obtained from a collection of laboratory test results as a lunction of shear strain amplitude No comprehensive study to measure P-wave damping has been conducted. -h-bgenerally-For the Al%UO it was assumed that P-wave and S-wase damping are the same. Tisseimiptionwael fotAPuWThe following study was performed to evaluate the effect of lower value P-wave damping,.
The sott-to-medium soil prolile with P-wave vehicity of 5Julit/see was analy/ed using SHAKE in two cases.
In Ca.se 1 the P-wave damping was assumed to be the same as the iterated S-wave damping tshown on Table 2A-10). This damping is in the range of 3 to 10 percent. In Case 2, the P-wave damping was computed using yo Op]4 <Y,Os
(
c where D, is the iterated value of S-wave damping.
The above relationship is based on /ero dissipation for volumetric changes. The values for ji, and V, were obtained trom Table 2A-10 and V was assumed to be Sju) ft/sec. It D was found to be less than 0.1 percent, e
the minimum of 0.1 percent was used. The response spectra in the tree-tield at the depth of 40 tt corresponding to the basemat elevation are compared in Figure 230.63-1. The resuhs are almost identical such that the curves for the two cases are essentially superimposed. The results are not atfected by the P-wave damping ratio. This is due to the fact that the soil column frequency at 40 ft depth is hrger than 31 Hert/ and the input motion at ground surface is ettectively retained at the basemat level in the iree-fiehl For this reason, use of smaller P-wave damping values is not expected to af fect SSI responses significantly.
b.
The P-wave velocity for each layer in the soil /mek protile was obtained from the strain-compatible shear wave velocity and the Poisson's ratios shown in Section 2A.4. The calculated P-wave vehicities are also shown in W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Hii 13 Response Revision 1 Tables 2A-9 through 2A-12. Depending on the SSI case and depth to the water table, the Poisson's ratios of the submerged layers were adjusted :f necessarv, to mai-tain the P-wave vehicity of the water (5JNN) II/see).
This adjustment et P-Aaw veh>:i'y and h*ce, tw Foisson's ra:ic, is necdej to telleet the Pavwe prol:.tpation specil in samrated media. The Poisson's ratio lor laprs above the water table are typical values appropriate lor each respective soil prolile. The SNI resuhs, particularly the horitontal responses, are believed to be insensitive to the change of Poisson's ratio. On the other hand, the certical responses fer each soil / rock case were governe i by the respectise shallow water table case due to the lact that use of P-wave v hicity of water results in less attenuadon of motion with depth, thus resulting in large effective foundation motion. The parametric SSI study on depth to the water table concluded that the water table at grade level is the governing condition for each respective generic soil prolile analy/cd. For these cases, the Poisson's ratio is assumed such that the P-wave velocity of the water is maintained.
SSAR Revision: NONE T
w 230.63(RI)-2 W Westinehouse o
NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 ACCELERATION RESPONSE SPECTRA 2.5 i
i i i iii i
i i i i i ii i
i i i iiii
- CASE 1 SOFT TO NED. STIFF SDIL
CASE 2 40 FT DEPTH H1 MOTION DANPING = 0.02 2.0 i
a z
O
[
- 1. 5
&W 4
0 4
A, 1.0 b
t in
.5 y
/
.0 10-1 10e 108 102 FREQUENCY-CPS Figure 230.63-1 Effect of I' Ene Damping W85tingh0USB
NRC REQUEST FOR ADDITIONAL INFORMATION
=r Question 230.66 11 Category 11 structures are adjacent to the nuclear island and will be drisen by the nuclear island, justity the adequacy of using Zone 2A requirements of the Uniform Building Code for the design.
Response
The only seismic Category 11 structure ath cent to the nuclear island is the high bay area of the radwaste buildine.
The analpis methods and design criteria for the seismic Category 11 structures are described in SS AR Subsection 4.7.2.f w hit h is being revised by the response to RAI 230.54. The design methodoloey has been revised to be based on the sale shut. lown earthquake instead of the Unilonn finilding C< ale.
Fhor respons 'pectra for the grade elevation of the nuclear island are show n in sheet I of Figures 2A-29,2A-30 and 2A 31. The spectra f or hard rock correspond to the ground input motion. The spectra f or the soil sites show little ampliliention abose the hant nick spectra. This indicates that the railwaste buildings will not be significantly driven by the nuclear island. llence design of the radwaste building to the SSE input without consideration of the effects of the nuclear island is adequate. Note that the specification of SSE input for design of the radwaste building is itsell conservative. As discussed in Regulatory Guide 1.143, the building could be desiened for lower seismic input.
SSAR Revision: NONE c
1 l
W Westinchouse o
NRC REQUEST FOR ADDITIONAL INFORMATION iiii = iiiig Question 230.68 Ciriity. in the SS AR, wl. ether concentric and du:d systems will be utili/ed when the Zone 3 requirements of the linitorm iludding Code are used for the design of the seismic Category 11 structures. It is the NRC stall's position that concentric and dual systems should not be utilized when the Zone 3 requirements are used for the design of the seismic Category 11 structures.
Response
The seismic Category 11 suuctures are the high bay area of the radwa3te building and the plant vent and stair tower.
The high bay area of the radwa<te building is a reinforced concrete shear wall structure. The plant sent and stair tower are concentrically braced steel structures. The analysis methods and design criteria lor the seismic Category 11 structures are described in SS AR Subsection 3.7.2.S. which is being revised by the regonse to RAI 230.54. The design methmlology has been revised to be based on the sale shutdown earthquake instead of the linitorm Buildmg Code.
SSAR Revision: None W Westinctic'Jse
=
=
NRC REQUEST FOR ADDITIONAL INFORMATION Question 230.73 Section 3.2.1.1.2 of the SSAR (pg 3.2-2) states that seismic Category 11 structures, systems and conyunents are designed so that the SSE does not cause mueteptable structural tailure of or interaction with the seismic Category 1 items. Section 3.7 of the SSAR qig. 3.7-1) also states that the seismic design of the AP600 seismic Categories I and 11 structures, systems and components is based on the SSE. Ilowever. Section 3.7.2.S (pp. 3.7-9) of the SS AR states that the seismic Category 11 structures me analy/ed and designed to prevent their collapse undes the SSE. and the scismic loads hir the design of these structures are analy/ed according to the 7.one 3 requirements of the l!nitorm Ilmhhng Code (l!IlC) usine an importance f actor of 1.0 Ilased on the above:
a.
clarily the im onsistency between the statements made in the SS AR.
b.
prmide the basis to demonstrate that the deripn of seismic Categcry 11 structuies that are hicated adjacent to the Ni structures, based on the Zone 3 requirements of the tillC with an importar.ce f actor of 1.0, will ensuie that the SSE will not cause unacceptable interaction with or tadure of any seismic Category I items.
Response
The seismic Category 11 stnictures are the high bay area of the radwaste building and the plant vent and stan tower.
The analysis metinids and design enteria for the seismie Category 11 structures are described in SSAR Subsection
?.7.2.8. w hich is being res ised by the tesponse to R AI 230.54. The design methodology has been revised to be based on the sat'e shutdow n earthquake instead of the linit'onn lluilding Code The annes and turbine buihling have been reclassified to non-seismie. SSAR Subsection 3.7.2.S. m resised by the response to RAI 230.54 provides the basis to demonstrate that the design of these non-seismic structures that are hvated adjacent to the nuclear island is s.ach that the SSE will not cause unacceptable interaction with, or failure 01, any seismic Category I items.
SSAR Revh. ion: NONE W WestinEhouse h
NRC REQUEST FOR ADDITIONAL INFORMATION Question 260.22 In Section 1.4 of the SSAR, " Identification of Agents and Contractors," Westinghouse identifies organizations that provide design and analysis support on die AP600 design. Discust the degree to which design certification engineenng and analysis activities are perfomied by the Westinghouse Electric Corporation. Provide details regarding:
a.
The extemal organizations that are being used to provide design, engineering, arvi analytical services, b.
The quality assurance provisions that have been invoked upon these extemal organizations, and The measures employed by the Westinghouse Electne Corporation to ensure the technical quahty of the c.
design, engineering, and analytical services procured from these extemal organizations.
Response
Westinghouse Electric Corporation has overall technical lead and program management responsibilities for AP600 design certification activities In addition Westinghouse directly performs engineering arx1 analysis activities in the following areas:
Reactor and nuclear systems arxi associated I&C, fuel harw!!ing, cooling water, and radwaste systems.
Reactor coolant pumps and control rod drive mechanisms.
Process instmmentation, protection, and control systems arul equipment.
Steam generators, pressurizer, reactor vessel intemais, fuel racks.
Nuclear fuel, control rods, and related core components.
Turbine generator and related systems.
Plant safety analysis.
Testing of safety systems atxi equipment.
To support and supplement the scope of work performed by Westinghouse, certain AP600 design engineering and analytical services are provided by extemal organizations. The work scope performed by extemal organizations is based on program requirements and the demonstrated expertise and experience of participating organizations.
a.
Extemal organizations perform engineering and analysis activitics in the following arcas:
Bechtel North American Power Corporation - A/E services for the Nuclear Islarki Buildings arxl associated systems.
Southern Company Services - A/E services for the Turbine Building and associated systems. Overall site plot plan development.
224 W Westinehouse
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f NRC REQUEST FOR ADDITIONAL INFORMATION ir "iii Ni T
Iturns & Roe Company - A/E services for the Annex Building, Diesel Generator Building and Radioactive Waste Building and associated systems.
MK.Ferguson Company - Constructability reviews, schedule and cost evaluatitas.
R.1,. Cloud and Associates - Piping, pipe support and snubber analysis.
Asondale Industries, Inc. - Modular construction engineering.
Chicago liridae & Iron Services, Inc. - Containment Vessel.
Columbia tinisersity - Fuel Assembly testing for Departure from Nucleate Boiling.
Orecon State University - Passive Core Cooling Integral Systems Test (low pn'ssure, quaner scale) liniversity of Tennessee - Reactor vessel lower plenum vortex suppression testing Pennsylvania State Ifniversity - Check valve reliabihty studies.
liniversity of Wisconsin - Condensation heat transfer experiments.
tiniversity of Western Ontario (Canada)- Containment coolmg wind tunnel testing.
SOPREN/ANSALDO (Italy) - Elements of NSSS systems and component design and testing, structural and layout design, piping analysis, safety analysis.
ENEI. (Italy) - Elements of safety analysis, PRA.
ENEA (Italy)- Automatic Depressurization System testing.
FI AT (Italy)-- Elements of NSSS equipment design.
1 SIET (Italy) - Passive Core Cooling Integral Systems Test, (high pressure, full bei ht)
F INITEC (Spain) Nuclear Island basemat, structural and layout design of auxiliary buildings, piping analysis, component specification.
ITI'E (Spain) - Elements of structural and layout design, piping design, component specifications.
ENSA (Spala)- Elements of NSSS equipment design.
ENtiSA (Spain) - Elements of nuclear design and safety analysis.
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l 260.22-2 3 Westinghouse j
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NRC REQUEST FOR ADDITIONAL INFORMATION i
f TECNATOM tSpain) - Elements of task analysis and operating procedures.
IIATAN (Indonesian - Elements of piping analysis, b.
Each external organization to which sesponsibility is assigned for some _nortion of the APMX) design for the Design Certification prognun is required to implement a quality assurance program that meets the requirements of ASME NQA-1 (ed. tion and addenda as specified in the APNX) Quality Assurance Program Plan). The extent to which NQA-1 applies is dependent on the nature and scope of tir work to be perfomied, and is subject to review and approval by Westinghouse.
The quahty assurance requirements for extemal organizatiotw that provide quality related services for which Westinghouse retains design rcsponsibility are commensurate with the nature, scope, and relative importance of the service provided. Depending on these factors, an NQA-1 quality nrogram may be required as described above, or Westinghouse may retain all responsibility for the work.
Individuals contracted by Westinghouse to perfomi quality related AP600 activities are required to work under the applicable Westinghouse procedures.
The pnmary method that Westinghouse uses to accept services (i.e. deliverables) from extemal organizations c.
to which design responsibihty is assigned is audit of their activities, as provided in NQA-1. In many cases, these organizations were selected as suppliers of engineering services because of their expenise in the assigned work. NQA-1 requires the supplier's quality program to provide for reviews of documents prior s release to assure that they are correct and complete. The supplier's intemal audit program provides oversight in this area.
Westinghouse audits of the suppliers further assure that appropriate quality programs are being effectively implemented. In addition, Westinghouse has established a program of independent, interdisciplinary design reviews of selected systems and design features.
Where Westinghouse retains design responsibility, Westinghouse verifies and approves the work in accordance with the Westinghouse QA program. Work performed by contracted individuals is also verified aryl approved in accordance with the Westinghouse QA program.
SSAR Revision: NONE PRA Revision: NONE 2m224 3 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION
_y Question 270.4 Section 3.11.1.1 of the SS AR states that a master list of :,afety-related electrical and mechanical equipment and a summary of electrical and mechanical equipment qualification results are maintained as part of the equipment quahfication file. The SSAR should identify w ho will he responsible for establishing and maiiaaining these filn
Response
Westinghouse will maintain the equipment qualification file to proside for review, audit, and inspection required for design certification of the APn00. TFe Combined License applicant must verify that the equipment qualification file is maintained during the equipment selection and procurement phase.
SSAR Revision:
l Revise the paragraph in Subsection 3.11.1.1 as follow s:
A comp!cte list of safety-related electrical and actise mechanical equipment that is essential to emergency reactor shutdow n. containment isolation, reactor core cooling, or containment and reactor heat removal or that is otherw ise essential in pres enting significant release of radioactive material to the environment is provided in Table 3.1l-l.
A master list of r.afety-related electrical and mechanical equipment and a summary of electrical and mechanical equipment qualification results are maintained as part of the equipment qualification file. The Combined I icense applicant must serify that the equipment qualification file is maintained during the equipment selection and procurement phase.
Add a line to Table 1.8-1 as follows:
Table 1.8-1 (Sheet 3 of 8)
Summary of APG00 Plant Interfaces With Remainder of Plunt item interface interface Ty pe Matching Section No.
Interface or Sub-item section 3,13 Maintain the equipment qualification file Requirement of Combined License 3.11.1.1 during the equipment selection and AP600 applicant program procurement phaw.
1 270.4-1 W Westinghouse l
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NRC REGJEST FOR ADDITIONAL INFORMATION maii9
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Question 270.5 hection 3.11.1.2 of the SSAR states that " Demonstration of quahfied life by test and analysis tor bothL with nocquate jt.stification, is pros.ded by equipment suppliers, incluump the efleets of aging w hen applicable." It is the NRC staff position that compliance with envirrumental qualification requirements is the responsibility the COL applicant, not the equipment suppliers. The SSAR should be corrected to clearly state w ho is responsible for compliance with requirements.
Response
The SSAR will be revised to explie! 'y state that the information is provided to the Combined 1.icense applicant.
Please also see the response to RAI 270.4 for a discussion of the maintenance of the equipment qualification file by the Cembined I.i::ense applicant.
SSAP. Revision:
Resise the ninth paragraph of Subsection 3.11.1.' as follows:
Safety-related systems and components exposed to a harsh environment have a qualified life goal of 60 y ears.
Demonstration of qualified life by test and anal) sis (or both), w ith adequate justification, is provided by equipment suppliers to the Combined License applicant, in.luding the effects of aging when applicable. For critical components susceptible to aging, a qualified life is established that includes the effects of the total integrated radiation Jose experienced at their respectis e locations within the plant. When a 60-year qualitied life is not achievable, a shorter qualified life is established, and a replacement program is implemented.
l 270.M 3 WestlnFhouse
i NRC REQUEST FOR ADDITIONAL INFORMATION Question 410.130 s
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l or the SWS. Section (1.2.1.1 of the SS AR states that tailure of the service water system or its components will not allect the ability of safety-related system to perform its mtended saf ety lunction. There is no similar statement in the SS AR lor the turbine building closed cooling water system.
Clarify it the above statement for the SWS is applicable for the turbine buihling closed cooling water system. Does the turbine building closed cooling water system meet the guidance of Position C.2 ot RG l.29 for non-safety-related portions? Provide the bases for this position.
response:
The statement made about the SWS in SSAR Subsection 9.2.1.1.1 is also true f or the turbine buihling closed cooling water systein. Fadure of the turbine buihling closed cooling water system will not atiect the ability of salety-related systems to perfonn their intended tunction. The inrbine building closed cooling water system meets the guitlance el Position C.2 ot Regulatory Guide 1.29 since system components are I icated away f rom or separated from salety-rel;ueil systems, structures or components such that the possibility of d.unage to satety-related components due to an SSE is eliminated.
SSAR Revision NONE 410.130-1 W WestinEhouse
NRC REQUEST FOR ADDITIONAL INFORMATION is Question 410.191 Are there penetrations in the walls between electrical equipment naims? How is thxid water in an electrical equipment area that may result Irotn firefighting actisities or fhwid water due to a crack in a fire protection system (FPS) water lint in the corridor of the electrical equipment areas of the auxiliary building prevented f rom spreading to othei nunns?
Response
Penetrations between nxims of dif ferent divisions will be sealed as required to maintain the lire rating of the barrier.
Those penetration seals below the maximum thod level in the compartment will also be watertight. Flood water originating in the corridor 01 the electrical areas flows under the stairwell door then down to elevation W 6" as described in SS AR Section 3.4. Some of this ihnxl water will spread under the doors into the adjacent electrical room s. This water wiis drain to the sump at elevation W -6" via o[vn floor drains. Separate drains are provided f or each Class IE electrical division and the non-Class iE electrical nxims. The maximum thod height in the Class lE electrie areas was conservatively calculated to be less than 3 inches. All salety-related electrical equipment that could be adversely allected by this lhxxl water will be hicated above the maximum thxwl height.
SSAR Revision: NONE W Westinahouse a
NRC REQUEST FOR ADDITIONAL INFORMATION Question 410.193 llow are battery rmms protected f rom water (both thxxl and spray) if the 1" deminerali/ed water system (DMWS) piping f ails in the ausiliary building corndor? Although the DMW lines are routed in the corridor, water spray can still allect the equipment in the battery rooms il the doors are not closed. Are there requirements for closure of these doors? Also, if these doors are not watertight, how is thuxhng in the corridors from a l'ailure of the FPS or DMWS piping prevented f rom affecting multiple battery rmms?
Response
The 1" deminerali/c hne is excluded f rom the pipe break analysis per SRP 31>.2, Revision 2. para. B.3.c.(l). In any case, the thu ahng and spray af fects are bounded by tire tighting activities w hich are addressed in SS AR Section 3.4.1. The battery nxun doors are normally closed and be ause the doors are fire rated, they have automatic closers.
These doors, however, are not watertight such that, it thxxhng were to occur in the corridor, the water is assumed to be uniformly distributed throughout level 1.
For this case, the maximum thiod level is calculated to ik approximately 6 inches (SSAR Section 3.4.1). This maximun. water lesel is substantially below the terminal height on the first row of batteries which is at 31 inches.
SSAR Revision: NONE 410.193-1 W Westmghouse
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NRC REQUEST FOR ADDITIONAL INFORMATION U ma 11 d'
Question 410.205 Identify those systems w hich are classified as moderate-energy systems based on the 2-percent and 1-percent rules.
Response
SSAR Table 3.6-1 provides a list of moderate-energy fluid systems that are considered for protection of essential systems. The piping lines from Table 3.6-1 that are classified as moderate-energy because these lines experience high energy conditions for only a short period of time are prosided below. The normal residual heat removal system lines, and the start up feedwater lines are classified as moderate-energy based on the 1 percent rule. These lines experience high energy conditions for less than i percent of the plant operating time. The spent fuel pit cooling sy stem is classified as moderate energy based on the 2 percent rule. This sy stem experiences high energy conditions for less than 2 percent of the system operating time.
SSAR Revision: NONii 410.205-1 3 Westinghouse
NRC REQUEs ODITION AL INFORMATION
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pr Question 410.207 Proside justification for why rotating components outside containment that are operated less than 2 percent of the time are also excluded as missile sources.
Response
The criteria that rotating components that are operated less than 2 percent of the time are excluded as missile sources provides for the exclusion of equipment that provides a minimal risk of missile generation. The motor for a motor operated valse is a typical example of equipment er.cluded by this criteria. This criteria has been previously
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approsed by the NRC.
SSAR Revisions: NONE 410.207-1 W Westinghouse
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NRC REQUEST FOR AFC!TIONAL INFORMATION Question 410 211 The March IX.19"3. response to Q410.52 states that protection of safety-related equipment trom turbine generator (TG) missiles is described in Section 3.5.1.3 of the SS AR, and is achiesed by proper orientation of the TG set and the use of f ully integral lowtressure turbine rotors. The report on rotors with shrunk-on discs was approsed by the NRC m Reference 2 of Section 3.5.1.3 and the methodology f or fully integral motors was submitted in Reference 1 of Section 3.5.1.3. These ref erences have been deleted f rom the SS AR. The stalt lwlieves that these references are appropriate and should be included in the SS AR. In addition. the March 18. 1993. response mentions a Reference 3, which th" statt cannot locate. Identity the reference and include that in the SS AR.
Response
The references hase been un luded in the SS AR as shown below.
SSAR Revision:
Revise second paragraph of Subsection 3.5.1.3 and add Section 3.5.5 as tollows:
The turbine and dise design is described in Section 10.2 Protecuen is provided by the orientation of the tuibine-generator and by the use of fully integral low-pressure %nme rotors. Analyses of the probability of the generation of missiles have been submitted to the NRC staff ine report for rotors with shrunk-on discs was approved by the NRC staf f (see Reference 1). The methodology mr fully integral rotors was submitted in Reference 2. Preliminary staff review (see Reference 3) agreed that the fully integnd low pressure rotors may be less susceptible to stress I
corrosion cracking than the shrunk-on dises. In the meeting on November 5.1992 between the NRC statf, EPRI, and turbine vendors, it was concluded that the turbine failures were not a safety issue and that the inspections recommended by the turbine vendors to ensure availabihty and reliability were more than suf ficient to ensure an acceptably low probability of missile generation.
3.5.5 References 1.
NRC Saftty Evaluation Report, letter from 11. D. Liaw to J. A. Martin dated December 27,1984.
2.
WSTG-4 P Proprietary and WSTG 4-NP, Non Proprietary. " Analysis of the Probability of the Generation of Missiles from Fully Inteeral Nuclear Low Pressure Turbiacs." October 1984.
l 4 Letter from John C. Tsao NRC to Daniel Fridsma Westinghouse PGTD, received 7/2/91, 1
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410.211-1 W WestinEhouse i
NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 410.212 is there any safety-related equipinent in the 25' strike zone of the turbine generator?
Response
As stated in SS AR Subsection 3.5.1.4, xdety-related stnNtures.systeins aad coniponents are heated outside the high-velocity, low-trajectory inissile 25 degree strike /one.
SSAR Revision: NONE l
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l 410.212-1 W85tingh0USB
NRC REQUEST FOR ADDITIONAL INFORMATION
=- =j Question 410.215 Are any safety-related systems or sy stems important-to-safety protected from missiles outside containment solely by providing sufficient distance between them and the missile source? If so, w hat is the minimum safe distance?
Response
'I here are no safety-related sy stems or components that have been identified as being protected from missile sources solely by providing sufficient distance between the sy stem or component and De missile sources. Spatial separation is acceptable for protection from missile sources under some conditions. Where spatial separation is used the range and trajectory of the generated missile must be shown to be less thu the distance to or directed away from the potential target. He Alt 00 has no equipment important to safety as defined in RAI 410.27. that is, nonsafety-related equipment whose failure could adversely affect the ability of si fety-related equipment to perform its safety function.
SSAR Revision:
Add an item to the list following the fifth paragraph of Section 3.5 as follow s:
Spatial separation may be used to demonstrate protection from missile hazards w hen it is shown that the range and trajectory of the generated missile is less than the distance to or is directed away from the potential target.
410.215-1 W Westinghouse I
NRC AEQUEST FOR ADDITIONAL INFORMATION A
Guestion 410.217 ProsiJe justification for the statement that rotating equipment in the ausiliary builJing is not a credible missile source if the equipment is used leu than 2 percent of the time. This includes pumps that operate < 0 percent of the time and nuitors for vahe,perators and methaniud handling equipment and pumps.
Response
'The criteria that rotatirig (oniponents that are operated leu than 2 percent of the tintie are excluded as missile sources prosides for the culusmo of equipment that prosides a minimal risk of rniwile generation. The motor for a motor operated s ab e is a ty pical example of equipment escluded b3 this criteria. This criteria has been presiously appros ed by the NRC.
I SSAR Revisions: NONii t
410.217 1 3 Westinghouse l
NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 410.218 What methmlology is used to determine whether a pump or motor casing can contain a missile generated by the failure of rotating equipment?
Rosponse:
See the response to R AI 410.55 for a discuwion of the methmlology used to demonstrate that housings (pump and motor casings) can contain miuiles generated by rotating equipment. See the Resision i to the response 'o R AI 251.1I for a discurion of the application of this methmlology to the containment of the reactor wolant pump.
SSAR Revision: NONii W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION A
Question 410.220 1he Marc h 18,1+0, response to Q410.67 states that rotating equipn.em in cmda;n:r-nt is climinated as a missile source fer one or more of the follow ing reasons:
lajuipment used < 2 percent of the time is not considerni a miuite sonr.e. This includes the reactor coolant a.
drain pumps, containment sump pumps and motors for s ah e operators and mechanical handling equipment and pumps b.
Pumps and fans such as the reactor casity supply fans are located in compartments surrounded by structural Z
concrete walls and contain no safety-related sy stems or eqaipment and so are not considered missile sources l(otating eqmpment with a housing or enclosure that would contain fragments of postulated impeller f ailure is c.
not considereti credible; d.
Non safety related rotating equipment in compartments w. i. safety-related equipment that da not proside other separation features hase design requirements for housings, r enclosures to retain f ragments from postulated failures of rotating elements.
Pioside justification for not considering equipment inside containment used less than 2 percent of the time to be erethble miuiles.
Response
The criteria that rotating components that are operated less than 2 percent of the time are excluded as missile sources prosides for the exclusion of equipment that prosides a minimal risk of missile generation. The motor for a motor operated vahe is a typical esample of equipment excluded by this criteria. 'this criteria has been presiously appros ed by the NR('.
SSAR Revision: N O N ii f
w 410.220-1 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION p
Question 410 222
'Ibc Marc h 18, lW3, response to Q410.64 states that no sources of priinary and credible secondary inissiles froin which safety-related equipment inside containtnent must be protected have been identified. A limited number of fans inay require design prosisions to confirm that they are not a missile source. Where is this discussed in the
%AR? Address this issue
Response
'the last sentent e of the second paragraph of Subsection 3.5.1.2.1.4 states *Nonsafety related rotating equipment in compartinents uiih safety-related systems or components that do not proside other separati.an features has design requitcmt nts for a housing or un enclosure to retain fragments f rom postulated failures of rotating elements.* The icquirernent for a enclosure or housing to contain f ragments of a postulated impeller f ailure can be esaluate;l once the fans are procured and the layout and module design is finalized for nonsafety related fans in the auxiliary building. 'this evaluation is the responsibility of the Combined license applicant.
SSAR Revision: NONI!
410.222-1 W ggy,m..
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NRC REQUEST FOR ADDITIONAL INFORMATION
'iii i j}:
Ouestion 410.224 Are any safety-related systerm or spiems itnportant to safet) protected from aniwiles inside contaitunent through the use of turriels? Il so, what are the barrier distierbions (wall thicknesses, etc.).
Responne:
Yes. As noted in Sub e(tion 4.5.l.2.1.4. salcty-related systems and components located inside the containment are protetled from miwiles generated m other ponions of the plant by the lhitt strustmal concrete shield building. The APfiOU has no equipment important to salcty as defined in RAI 410.27, that is, nonsalety-related equipinent w hose f ailure conhl adversely allect the ability of safety-related equipinent to perform its salcty f unction. Also, safety-related systems and components located inside containment are protected Inim iniwles reneraied inside containment by the interior structural conc rete. The wall and root of the stueld building are % and 24 inches lhick respectisely.
The interior structural concrete vanes m thickness. In the lower portion of the containment tiu equipment is hicated m cavities in the inass concrete with many leet of concrete between the (avities. At higher elevations the interior structural concrete sinnettares are the walls of the ref uelitip canal arid the shiehl walls around colliponefits al lite operating tiet L and maintenant e floor levels. The thit kness of the ref ueling canal walls is 48 inches and of the stearn rencrator compalment walls is 30 inches. These walls are adequate to protect against internally generated mimles.
The protedores used to determine that the lurrier design is adequate are outlined in Subsection 3.5.3.
SSAR Revision: NONE l
l W WestinEhouse
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NRC REQUEST FOR ADDITIONAL INFORMATION r'~
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Ouestion 410.225 Identify safety-related equipment and equipment important-to-safety that are sub.ieet to missiles f rom non-seismic Category i structures, sptems, and components iriside containment, and discuss how this equipment will be protected (discuss non-safety related sptcms in.,ide containment with regard to their potential for generating missiles w hich could damage safety-related equipment).
Response
Subsection 3.5.1.2.1.4 prosidn the evaluation of the potential for missiles from nonsafety-related spiems and components inside containment. See the response to R Al 410.222 for additional diwussion on the need for design requirements for nonsafety-related f ans.
SSAR Revision: NON!!
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VJ Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION I
Question 410.22G Are any safety-related sptems or sptems important to-safety protected from missiles inside containment miely by providing suf ficient distance betw een them and the minile murce? If so, u hat is the sninimum safe distance?
Response
There are no safetprelated systems or components located inside containtnent that huse been identified as being protected from missile sources solely by providing sufficient distance between the system or component and the missile sources Spacial separation is acceptable for protection from minile wurces under some conditions. Where spacial separation is used the range and trajectory of the generated missile must be shown to be less than the distance to or directed away from the potential target. The APtu) has no equipment important to safety as defined in Q410.27, that is, non-safety-related equipment whose failure could adscrsely affeet the ability of safety related equipment to perform its safety function.
SSAR Revision:
See the response to ital 410.215 for the SSAR resision related to this question.
410.226-1 W Westinghouse l
NRC REOUEST FOR ADDITIONAL INFORMATION Ouestion 410.??7 Ilow w di the nintrol rooni bc stotnInt f roni inissiles penciateil inside containtnent.'
Response
The nuun control sooni k located in the auxiliar) buihhng. Area
- outside the n ntaintuent are protectni f roin missiles generatni niside containtoent by the containinent shell and shield buihling. Additionally areas inside the ausihary buil kne are protected froin uiternally generated iniviles outside the ausiliary buihline by the structural uincrete u nits, roof, and floors of the auxiliary building.
SSAR Revision: NONi; 410.227 1 W WestinEhouse
N6C REQUEST FOR ADDITIONAL INFORMATION Question 410.228 Include i list in the SSAR of the systems that must be protected from miuiles generated by natural phenomena.
Response
As noted in Subsc: tion 3.5 2. the systems required to be protected from externally generated missiles are the systems required for safe shutdow n.
'llie requirements for safe shutdown are discuued in Section 7.4 with the spetific list of sy stems required m Subsection 7.4.2. 'lhe spent fuel pit also must be protected. The areas to be protec ted are outlined in Subsection 3.5.2. These areas are within the nuclear island structures (containment shield building and auxiliary building).
SSAR Revision: N O N i!
410.228-1 W Westinghouse
NRC R'.OUEST FOR ADDITIONAL INFORMATION Ount, tion 410 220
'i he N1.ut h lx. UN i, response io 04111.70 states that the estonateil prohahint) ol win,l specils ricales than the h Ni soph 1)ll'l is hetw een 10' a'i I 10 pel year for the \\PfitHhlesign at a uotsi hration anyw here ni the t Onternons l '.N
' lins shouhl le ins lutled in the appsopnate set tion of the ';s Alt.
Henponw
'lhe piolubilits has been intin leil in the $5 AR as show n Irlow.
SSAR Revision Reuse Subset tion 4. U.I as follow s:
't he ilcsiro pai;uncters apphcable to the ilesiyn hash toenailo.ne as follow s:
hiatunum wind specil HN1 inph N1asinnun totational specil - 240 mph hiaininnn translational speed - to niph Ra(hus of inasinnnn totational wind lioin center of totaado - 150 It Annosphenc pecuore ilnip - 2.0 psi Rate ol' prewnie ( hanre 1.2 pstAcc e
11 is estiinated that the prohahility of wital speeds picater than the 3tWl-mph design basis tontado is betwwn 10* anil 10 ' per year for an APNiti at a worst hication anywhere within the motirnons llniteil Staics.
W WestinEhouse
NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 410.23C liow will the tonnol rooni be protected Iroin niiwles rencrateil by natural phenomena?
Renoonse:
't he ruain control ronin is located in the auuhary bushhng. Areas inside the ausili.:!y bioltling are protecteti trom nussiles rencrated outside the ausihary buihhng by the structural concrete walls roof, and 11oors of the ausihary buihhny. The walls sunounthny the snaiti wntrol rooni are reinforced mot rete structural walls.
SSAR Revision: NONii W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Question 410 231 Ptoude.potiin, anon for the use of 2 psi pressure drop rather th.ui the 2.25 psi preuure drop pes ilied in Regulatory Gunie IRG) 1.76.
Response
The 2 psi prewure drop is t onsistent w sth the s[rcification of the 3tu roph tornailo. This value was approved by the NR(' stall as doconiented in the Dra!I i uial Sately livahution Report (page 1.4 22) on the Al.WR Utahty Retpurenients I)octunent for pawne plants. hsued August 41, l'193. It is espetled th.it this position will be in.untained in the i inal Salcty 1: valuation Report to be iwned in IW4.
W WestinEhouse
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NRC REQUEST FOR ADDITIONAL INFORMATION gr Question 440.75 Section 5.2.2.1 of the SSAR states th.it the siring of the pn ssurver safety valves for over-pressure protection of the RCS dunny pou cr operation and transients is based on die analysis of a complete loss of steam flow to the turbine, with the reacto operating at 102 percent of rated power.
'The acceptance critena in Section 5.2 of the SRP states that the safety valves should be designed with a.
sufficient margm available to a.munt for uncertainties in the design and operation of the plant, assuming that the reactor scram is initiated by the secorxl safety grade signal from the reactor protecuon system.
IIase these been included in the analysis?
b Table 5 417 of the SSAR provides the lesien parameters for the pressuiizer safety valve. What are the uncertainties of the set pressurvs of the pressuii7er safety vaive?
c.
Piovide die result of the analysis, mcluding the sequence of events of the design basis incidents for sir.ing the pressuriter safety valves.
Response
a.
The safety valves provide margin to account fer uncertainties. Reactor inp signals were not explicitly relied upon in the sidng of the pressurves,afety valves. See additional discussions under the response to past c, below.
Uncertainties explicitly allow ed for in the design aod operation of the plant in the sizing of the pressurizer safety vahes are as follows.
2% uncertainty in nuclear power 4'F uncertainty in reactor vessel T,,,
10% nacertainty in steam generator secondary sids P9d mass Steam generator UA based on 10% of the steam ganerator tubes plugged Minimum available pressurizer steam space bawd on statistical combination of control and protedion system uncertainties No credit for steam generator secondary side PORV's No relief through steam generator secondary si'le saf ety valves until steara generator secondary side pressure reacles 103% of steam geiwratos shell desinn pressure.
Steam generator secondary side safety valve relief capacity at 103% of steam generator shell design pressure was taken to be no greater than the plant sated s. team flowrate No credit for plant control systems No cre ht for powef reduction due to reacn"ity feedback b.
As specified in the AP600 Technical Specifications (T.S. 3.4 6) the pressurizer safety valve setpoint tolerance is 11% of set pressure.
440'75-1 W Westinoh0use u
NRC REQUEST FOR ADDITIONAL INFORMATION A
The pressunxr safety vahe.s are sired to carry the maximum pressuriier volumetric insurge following a c.
cornplete loss of 1.ul arxl feedwater from !'12% of rated power, Following the loss of load and feedwater, reactor power is maintained constant at 102% of rated power. No credit is taken for reador inp or reactivity feedbact during the transient. Credit is taken for actuation of the sicain renerator sewrulary side safety valves, which are rnodeled to pass plant rated steam flow at 103r?c of steam generator shell design pressure. The pressuri/er safety valves are modeled to pass a large wiumetric stcarn llow (for example 1(K) ft'/sec) at 103%
of reactor wolant sysicm design pressure ipressurver safety valve set pressure plus 39 accumulatioro so as to negate compressibihty cliects durmg the sizing calculation For the limitmg transient evaluated, the loss of load armi feedwater was simulated to <>cciir at 10 seconds.
Prirnary and secordary teraperatures then increased and the steam generator secondary side safety valves opened at 24 seconds into the transient. Tie preuuriter insurge rate peated at this time at 25.7 ft'/sec. The peal thscharge rate for the steam generator secondary side safety valves occurred at the time of ternunation of the transient simulatior'. 30 seconds which corresponds to 20 seconds after initiation of the loss of load and feed, and (bd not t seced 85% of plant rated steam flow This is well below tie actual capacity of the steam generator secorxlary side safety valves.
The specific soluine of saturated stearn at 2575 psia (1039 of RCS design pressure)is 0.12345 f t'/lbm.
(25.70 ft'/sec) * (3600 sec/hr) / 0.12346 ft /lbru = 749,393 lbm/hr 2
The required valve capacity was then conservati$ely established at 750JK)0 lbm/hr at 2575 psia.
SSAR Revision: NONE i
440.75-2 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Question 440.77 Table 5.4-17 of the SSAR lists the ensironmental conditions for the pressurizer safety sahes and the relief sabe in the nottual residual heat rernoval systern (NRHRS) suction line to be 50 to 120"l:(arnbient teinperature) and to be at a relatise hutnidity of 0 to 100 percent. Does this arnbient temperature condition enselop the accident conditions?
Response
The ensironmrnta! conditions provided are the normal ensironmental conditions for the specified vahes. The abnormal operating ensironment for these vah es is pros ided in SSAR Appendix 31), Table 31).5 3. These abnormal ensironmental conditions enclope the accident conditions for the Al'600 I
SSAR Revision: NONii i
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440.77-1 W Westinghouse l
NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 440.82 Action A of Technical SpeciGcation Ifo 3.4.13 requires that, if an retunulator is not isolated w hen the accumulator pres.ute is greater than or equal to the inasimum RCS preuure for the esisting cold leg teinperature a
allow ed in the preuure-ternperature limit report (I'll_R), the aficeted accumulator inust be isolated u ithin one hour.
It' not Action 11,1 requires that the RCS cold leg temperature be increased to a level acceptable for the esisting accumulator pressure allow ed in the Irl'IR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. liowever, tro 3 4.13 prohibitutarting the RCS pump when the RCS temperature is greater than 2(M)"I', if the preuuriter level is greater than or equal to 92 percern, What is the means of increasing the RCS temperature?
Response
The reactor coolant pumps would be used to increase the temperature of the reactor coolant sy stem. If the lesel in the preuuriter is greater than 424 and the reactor coolant 3 stem teniperature is greater than 2(x)*l, then the preuutiter level would be reduced to below 924 prior to starting the reactor coolant putnps. If core deca) heat is inailable, the reactor coolant system temperature could also oc inercased by reducitig restor coolant sy stein shutdow n cooling sia the residual heat removal heat eschangers.
SSAR Revision: NONii l
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440.82-1 i
3 Westinghouse l
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NRC REQUEST FOR ADDITIONAL INFORMATION Question 440.84 Position 116 in the liranch Techmeal Position lllTP) RSil 5-2 states that the !!!'OP sy stem should meet the requirements of Reguhitory Guide (RU) 1.26 on qu dity group classitkations and standards for the components of nuclear pow er phmt*,.
In Revision I to WCAP. 3054. Westinghouse takes exception to some of the RG 1.26 guidelines for the ll LOP system because of the wa) in w hich the APMM) provides safety-related functiens. Provide a detailed discussion on these exceptions :.od proside justification for the acceptability of these exceptions.
Response
Regulatory Guide 1.26 describes a method for determining acceptable quality standards (i.e.. Quality Group 11. C.
or in for the safety related >> st, ms, structures, and componerits containing radioactive material, w ater. or steam that are not part of the reactor coolant pressure boundary as defined by 10 (TR Part 50 paragraph 50.2v. or specifkally excluded from the requirernents of 10 CI'R Part 50 paragraph 50.55a (c)(l).
l or the Al'600, 'he normal residual heat removal pump suction relief s alte performs the function of reactor coolant system low temperature overpressure protection. This component is located outside of the reactor coolant pressure boundary as defined by 50 2v and is assigned to Regulatory Guide 1.26 Quality Group 11 because it serses as a containment isolation vahe. Regulatory Guide 1.26 does not state the Quality Group that is to apply to the low temperature oserpressure protection function. Designing the normal residual heat remos al pump suction relief s ah e to Regulatory Guide 1.26 Quality Group 11 requirements meets position 11-6 of liranch Technical Position RSil 5 2 and Regulatory Guide 1.26.
WCAl'-13054. Rev. I will be revised to indiente conformance with pos; tion 11-6 of IITI' RSil 5-2.
SSAR Revision: NONil I
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e NRC REQUEST FOR ADDITIONAL INFORMATION qp-nn i
m Question 440 86 Sec tion n.4 of the SS AR staics that the ernerrency t ore heat scinmal lunt hon of the PNS spteni is available at seat tot tuotant s3 sicm t on hhons mt knhny shuhlow n and ref ueline. Is (los stateinent true f or eclueloir conthlions w ben snost of the IRWST water n transf erred to the reluchny lnot and not avadable lor the PNS sy sieni operation.'
Responso A paune saf ety-sciated feature a asailable durme all modes of plant operat on intluihny shutdow n.
Doring shuhlow n operatoms with the sca,for ( oolant pressure boundary is intact, the pawise resulual heat removal systern heat cu banper is asadable. With the reactos coolant system prewure bounder) open, automatic deprewuri/ation y sicm senting and in-contauunent ref ueling water stolare tank inletlion aic asa lable. While the tractor coolant sNem n in the reluchng snode, the llooded telueliny cavity pim iiles deca 3 heat removal capabihties. These qstenn provnic salct3 related f ailure tolerant means of icmm my decay heat should the nonaalctyaciated nonnal conhny spicm 1;nl.
In (1inthation of the staternent matie in SS AR section 6.1, the pawise (ole cmhny splein (hies not duet fly provide the safets-Iclated tucans of decay heat removal dutmp scinchny ope:ations. The PNS heat reinoval features are not tequued to be aunlable durmy reluchny opciations when inost of the IRWST water has been transicoed to the reluchny un ity. The reluchnr cavity ponides sigmlicant heat removal capabihty. Awuming that the RNS lads just uhen tht ref ueling cavits n flooded. it would take about u hours to beat the icluchnp t auty water to boiling and about 6 day to herin to uncm cr the f uel. This pnnides ample tune to (lme the contaimnent of to secure alternate water supphe%
! or lutther thstumon in scrants to t ore det ay heat semmal during icluchny operations, reles to SS AR scchon 6.t 4.4.4. ! oss of Normal Residual lleat Remmal (i.olmy Duriny Reluchng.
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64 PASSIVE CORii C(X)l.ING Si STEM I meirency (ore decay heat seinmal Pnn nie (ore det av heat removal dunng transients, accalents or w henever :he nonnal heat removal paths are lost.
This heol icmoval f uochon is available at reactor wolant sysicm conthiions including shutdowns and reluchny.
l During relueling operations, w hen the IRWST is drained into the ref ueling cavity, other passive incans of core decay heat removal are utili/ed. Subsechon 6.3.14.4 provides a description el how this is accomphshed.
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440.86-1 W Westinghouse 1
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NRC REQUEST FOR ADDITIONAL INFORMATION l
Question 440.89 Sechon 6.'4.1.2 of the SS AR states tius the itequens) of autonune ileptenumatnin systein actiution is lunited to a low piolubihty to redute s.delv risks and to luinunUc plant vulare. I) cline the teHn " low proluhiht)." and esplain how ilus roal n acinesed.
Response'
'lle ecsponse to R AI 471.20 pimides an estunaie of the Ircquency of itutbertent A1)S f or the AP600 The resula inibcate that the hequency of inaiheitent AI)S at tudion is approtituatcl) 2ii-4/3 r. Tha f requens y h sunilar to the hequency of very sinall and sinall 11)('As used in the APNNi PR A (1.0715 4/)
).
Alues uir a " low probabihty" of uush eltent A11S as tualnin is as hieveil throuyh the use of certain design features anni vahe opening charat terhtics. 'these leatures and charattenstics inchale the loilowiny tako included in the response to R AI 471.20):
1.
The SS AR chapter is nah sn show s that A!)S operanon does not otcut diump non-l.()('A events or steam a
rencrator tube sup!ures. It n only espected to occur dunny an ituihcrtent Al)S event.1.O('A's. or long terin sale shutdown when all ac power is lost for inore than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
Me Al)S is acliuted with 2 out of 4 lorie. As a result, inultiple instiumentation ladures are required to inadsettentl> aituate the Al)S.
4 liach Al)S stare 1,2 and I line has two norinally closed sahes ili series. The Al)S stare lour vahes are mtcihk Led lo prevent their opening at nonnal R('S piewures. As a tesult, inultiple tailures are net ewary to inativentantly actuate the AI)S.
4 ll A1)S is actuated in.uheltently of 101 a sinall R('S L()('A. the jourth starc will not he actuated awuunny that the operators stant the normal seudual heat removal pumps. These pumps pimide the R('S injection tunesion and cause the ('MT injet : ion to stop with the ('MT lesel abiwe the f ourth stare at tuation wipoint.
5.
Il the nonnal residual heat iemoval. system is unavailable in an inaiheitent AI)S event and the Al)S lourth stare vahes opet.. the containnt;et will slowly thnl to its maximum level mer several days. Recovel) of the nonnal residual heat rennnal system dur;ng this time allows the floodup to be terminated.
SS AR Revhion: Nt )Nii W Westinghouse
e NRC REQUEST FOR ADDITIONAL INFORMATION
!!E
"!!!m Question 440.91 Sectnen (i.C'.1,1 of the SS AR staics tlut the PRilR heat cu taneers are uinnected to the R('S through a corninon inlet line Isoin one RUS hot ler (throuyh a tee hoin one of the lourth stare ADS hnes) atul a uniunon outlet line to lhe awwiated stearn rencrator cold ley plenuin, llecause a uinunon nunle f ailure (e.y.. a break of the conninon line4 wouhl ihsable both PRilR heat en hangers. <hscuss the reasons for this arrangenient.
Response
The pmise ecsidual heat removal piping is APWicquilunent slass A because it is part of the scactor oiolant sysicm pressure boundar). A breal of this hoe would be a 1.(X'A. for whn h operation of the pawise residual heat reinoval heat en hanger is not required for snscessiul mitiyalion of this L(K'A. In aihhtion. awominy a break of this hne dusiny a non-llX'A es ent w here the pawive residual heat removal heat vu hanper is requiteti to operate, is beyonti the destyn basis of the plant. In such a non l ora esent. it is init necewary to awume that a independent rupture of a reactor coolant pipe II (X'A) occurs.
SSAR Revision' N(INI!
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NRC REQUEST FOR ADDITIONAL INFORMATION
_.. q Ouestion 440.99 hheet I of Table 6.31 of the hSAlt for Case 1 "Non-LOCA, CMT Operation in Water Circulation Mode "
intlicates that the fluid for the direct s essel injection line (laation 35)is air. This appears to be inwrrect the fluid be water?
. Should
Response
Yes, the fluid should be w ater.
SSAR Resision-Resise Table 6.31, at Location 35, as follow s:
Table 6.31 (sheet 1 of 14)
AP600 PXS Process Flow Table (Safety injection)
Ca.e I: Nont OCA. CMT Operation in Water Circulation hiode I
location Description I'luid l'ressure feiop.
I' low Volume t psig)
( F)
(Ib/sec)
(f t3) l Ptr steam Steam 2,400 663 450 2
Ptr liquid Water 2,4(M) 663 ESO 5
Cold leg Water 2,400 550 Nat Cire 6
110t leg Water 2,4(X) o00 Nat Cite 10 Per'CMT line Steam 2,4(K) 663 0
1I ClXMT line Water 2,400 550 37 12 CMT inlet line Water 2,400 550 37 15 CMT ttop)
Water 2,400 550 100 16 CMT Water 2.400 120 1,WK) 18 FMT outlet line Water 2,400 120 49 20 Accum gas Nitrogen 700 120 300 21 Accum water Water 700 120 1,700 W Westinghouse 440.90-1 N
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NRC REQUEST FOR ADDITIONAL INFORMATION
'A 22 Accutti outlet line Water 2.400 120 0
25 IRWS) \\etil Air 0
120 0
2t>
IRWST gas Air 0
120 5 (NN) 28 IRWST unter Water 12 120 70,MN)
.40 1RWSI outlet line Water 12 120 0
32 cont sump cl valve Watet 0
120 0
11 Cont Mimp \\10Y Air 0
120 0
15 l)VI line Aif Water 2.4(N) 120 49 440.99-2 W Westinghouse
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NRC REQUEST FOR ADDITIONAL INFORMATION
!!Y Ouestion 440.102 Sheet 3 of 'lable hM of the SS AR. " Failure Males and Ellects Atulysis for PXS Active Cornponents." indicat s llut spurious opening of the acconiulator rutrogen supply / vent salves las no alety-related ellect since cash valve has enhet a nonnally-closed, redundant series isolation SOV of a (het t valve in eat h vent flow path tlut presents au urnul.nor nitrogen troni leaking out of the accumulator, w hich could preclude accuniulator injection. Clarily or proude a inore detailed PAID ' hat show s such a valve arrangesnent.
Response
11 cither PXS accumulator nitrogen supply valve (Y012 Allo should spuriously open, there are a liinited nuinhe of leakage paths that the Nitrogen could tale. Tlie leakage llow to any of these possible luths wouhl be bhicked by one of the following salves:
1.ine To:
1.ine #
Isolation Valve Containtnen: penttration Liu4 PNS Vo43 Vent to wnlaintuent l_ t M9 PNS V(45 Prewule relief to containnient IJMI PXS V(44 pil Tank Lt 47 PXS V3tvi MCDT LOSO WLS V(Wi3 These lines / valves are shown on the PXS PAID (SS AR Figure 6.31 eteept for the pressure regulator valve on the hne to the RCDT. Tiut vahe is shown on the WLS PAID (SS AR Figuie 11.2).
SS AR Revision: NONE W WestinEhouse
NRC REQUEST FOR ADDITIONAL INFORMATION n.i.i.
Question 440.107 During an iiudvertent opening of a steam pencrator reliet or salcty valve and ste;un systein pipe lailure, the teact(n is inlyied and the core inateup tant is astuated to inject borated water by the saf eguanis actuation signal.
Settions 6.4.3.1.; and 6.3.4.1.2 of the SS All luth state that, ahhough the lurated water solution does not provide sullicient negative reactivity to inaintain the reactoi suberitical the core is ultiniately shut dow n by the lurated watet solution. Explain how the reactor is shutdown by a lmron solution that does not provide sulticient negative reactivit) to inaintain the reactor subs ntkal.
Response
Plant coohlown irutially add,, positive reactiuty laster than the ChlTts) add negative reactivity timrated wateth lhiweser, as the couldown slows and eventually stops. so does the addiuon of positive reactisity. hicanw hile Imration continues and eventually compensates for the tusitive reactisity added during the cooklown and completel) shuts down the reactor. Itefer to analysis of itCS couldow n events in SSAlt secuon 15.1.
SS Alt iteve, ion; llevise SSAlt section 6.3.L1.1 as hdlows Ahhouyh the bosaied water uduiion loe not p. ovide.u fhient neyotive sem tivity tomaimain 4heaem tot subesitkal; the sese-is ukimately shui down by ths%.dution unil-The trip of the reactor initially brings the reactor sub critical.
The rapid ItCS cool down may result in the reactor returning to critical because the rate of positive reactivity addition (ItCS temperatute reductiou) exceeds the rate 01 negative reactivity addition thonin from the Ch1T). As the event continues, the ItCS couldown will slow down such that the continued Ch1T boration will return the reactor sub-etitical. I'the departure f rom nucleate boiling design basis is inet, thereby preventing f uel (Lunage.
lievise SS Ait
.in 6.3.3.1.2 as f ollows:
The negative nwiivity provLk d by49ienation4,1 the4oreanateup iank+h no +utsidimi io14 event ih+nwoor 4non seinoting to-esiticahiy41uring theiransienic The trip of the reactor initially brings the reactor sub-critical. The rapid itCS cool down may result in the reactor returning to critical because she rate of positive reactivity *idduion (ItCS temperature reduction) exceeds the rate of negative reactivity addition (boron from the Ch1T). As the event continues, the itCS cooldown will slow down sue., that the continued Ch1T 1mration will return the reactor sub-eritical. Ilowever; the-iore is shui 4Liwn-bv-the-hirated-wales-*,4utioncond i The departure Inim nucleate Imiling design basis is met.
440.107 1 W West.irlEhouse
NRC REQUEST FOR ADDITIONAL INFORMATION
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Question 440.114 Section 6.3.3.4.4 of the SSAR states that in the event of a loss of the NRHRS during refueling, if the containment is not scaled, imiling will reduce the refueling cavity water les el to the top of the fuel in about 5 day 5, but continued core cooling can be easily maintained by one of several methods, such as closing the containment and using multiple non-safety.related sy stems.
llecause there is no Technical Specification control for the non-safety sy stems (e.g., CVCS), w hat is the basis or justification to assume their asailability?
Response
SSAR subsection 6.3,3.4.4 provides a description of the operation of the passive core cooling system for postulated es ents that could occur during shutdown. As is stated in the SSAR, tne loss of the normal residual heat removal system during refueling would result in the refueling cavity water heating t.p to saturation in about nine hours.
Subsequent boiling would reduce the water level to the top of the fuel assemblies in about the days, if the containment is not sealed. The safety-related means to provide core cooling following this event is makeup via safety-related connec uons in the normal residual heat removal system frort u ernerg-r.ey, temporary of f-site saurce that would be brought to the site within 3 days. Availability of the emeigency, temporary of fsite source is identnied as Combined 1.ieense applicant responsibility in SSAR Table 1.8. SSAR subsection 6 3.3.4 describes the first line of defense for makeup, w hich is the nonsafety-related chemical and volume control sy stem makeup pumps. SSAR section 6.3.4 describes the post-72 hour safety related actiur:3 required.
SSAR Revision: NONii 440.114-1 W WO5tingh00S8 1
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NRC REQUEST FOR ADDITIONAL INFORMATION
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Question 440.115 Sections 6.3.7.7 arnt 7.4 of the SS AR state llut there is a inner that autoinatically attuates the autonutic deptessurt/ation systein il the ollsite and onsite ac power sources have been lost for about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, that the dc baricetes that power the ADS valus provide power for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and that the ADS valves will be opened bef ore the batteries are discharged. Proside the iletailed sequencing and timing of the actuation of various ADS stare sal.cs alter a loss of the ac power source.
Response
The f unctional diagrams of the Aim reactor protection system show the automatic depressuri/ation system actuation logic in the AP600 SS AR on figure 7,21, sheet 15. As indicated. there is a timer w h:.h would actuate autonutic depressurifalion system stage 1 af ter a sust.ned loss of ac power. The setpoint for the tuner has not been i
finali/cd but is anticipated to be apprmimately 23 l. ors. The other stages of the automatie depressuri/ation system would lollow based on their nonnal sienals. With.ae automatic depressuri/ation systern actuation t hanges discussed in the design change patLace subtnitted to the NRC an Feb. 15,1994, stages 2/3 are actuated a l'ised tirne alter stare I actuation. Stage 4 follow s stage 1/2/3 w hen the core makeup tanks drain to a low 2 level. The actuation of all of the autenutie depressurization system stages is estnnated to take less than 30 ininutes alter stage 1 actaation.
SSAR Revis:on: NON!!
440.115-1 W Westinghous3 i
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NRC REQUEST FOR ADDITIONAL INFORMATION m
l Question 440.127 Section 5.4.7.1.2 of the SSAR states that the reliability of the non-safety grade normal residual heat remual system (RNS)is achieved by using highly reliable and redundant equipment and a simplified system design, and the system piping and components are Safety Class C. seismie Category I for pressure retention purposes only.
Clarify or define the term " highly reliable equipment."
a.
i b.
Regulatory (iuide 1.29. C. I.d. requires that >> stems or portions of sy stems that are required for reactor shutdow n, residual heat removal or cooling the spent fuel storage pool are designated as seismic Category I and shall be designed to withstand the etfects of safe shutdown earthquake and e1nain functional. What t
are the bases for Jesigning the RNS piping and components as Safety Class C, seismic Categors I for pressure retentiortpurposes onh?
Response
lhe term
- highly-reliable" is nebulous and will be deleted as show n in the response to R AI 210.37. The intent a.
is to proside reliable equipment as spe,itied by its equipment ciassification in Taue 3.2-3 of the SSAR.
b.
'lhe conformance of the AINM) design with Regulatory Ciuide 1.29 is presented in Apptndix 1 A of Set tion 1.9 of the AlWM) SSAR. The pas,ise residual heat removal subs) stem of the passive core cooling sy u - n the safety -related means of : ore tesidual heat removal and is classified as Seismie Category I in accordan<e with the APMM) Classification System and Regulatory Guide 1.29 for this f unction.
'lhe portions of the normal residual heat remosal system that serve as a part of the reactor coolant pressure boundary or as a part of the containment isolation sy stem are designed to withstand the etfects of earthquakes and remain functional and are designated as Seismic Category I in accordance with the APNK) Classification Sy stem and Regulatory Guide 1.29 Tht remaining portion of the system located outside containment is classified as Seismic Category I by the APMX) Classification System, not to protect its capabilities to remove residual heat, but to protect the integrity of the subject pressure boundary following a safe shutdown earthquake, l'orther clarification is prosided in the responses to RAl's 210.37 and 210.61.
SSAR Revision:
Sections 1.9.5.1 and 5.4.7 will be updated to clarify the basis for classifying portions of the normal residual heat remos al sy stem as Al%0() equipment class C and Seismic Category I. The revisions are provided w ith the responses to RAl's 210.061 and 210.37.
440.127-1 W Westinghouse
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NRC REQUEST FOR ADDITIONAL INFORMATION l
Em m a Question 440.129 Section 5 4.7.1.2.3 of the SSAR states that the RNS is designed to provide cooling for the in-containment refueling w ater storage tank (IRWST) during operation of the passis e RHR heat -xchanger or durin ; normal plant operations w hen required to 'imit the IRWST water temperature to less than 212 "F during extended operation of the PRHRS.
Because ti-RNS is required to support the operation of the safety grade PRHRS and IRWST, w hat a.
are the bases for not requiring it to remain operational after a shutdow n earthquake?
b.
I igure 5.4 6 of the SSAR shows that the motor-operated valve in the IRWST discharge to the RNS suction line is normally open, w hich should be normally closed as show n in Figure 5.4-7. Correct the inconsistency.
Response.
The nornud residual heat removal system cooling of the in-containment refueling u ater storage tank (IRWST) a.
is not a safety-related function. The normal residual heat remosal s) stem is not required to remain operational following a shutdow n carthquake. Normal residual heat removal sptem cooling of the IRWST is not required for the IRWST or the passive residual heat removal system to perform their required functions to mitigate design basis events as analyzed in the Chapter 15 Safety Analpes.
Atter appro.imately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of passis e residual heat remos al heat exchanger operation following a doign basis event, the IRWST may begin to boil and begin to steam to the contaimnent atmosphere, long-term passive residual heat remosal heat exchanger operation results in prolonged IRWST steaming which leads to eventual actuation of the passise containment cooling sy stem. IRWST steaming to containment requires containment cleanup operations. If the normal residual heat remos al sy stem is available, manual initiation of its tw o trains before the IRWST begins to boil permits the normal residual heat removal sptem to maintain the IRWST temperature less than 212"F and serves as investment protection against containment cleanup follov hg a design basis esent due m IRWST steaming.
Intermittent cooling of the IRWST by the normal residual heat removal sy stem during normal plant operations ensures that the IRWST temperature is within the maximum initial temperature assumed in the Chapter 15 Safety Analyses -120'F as required by the plant Technical Specification Surveillance Requirement 3.5.4.3.
b.
Figure 5.4-6 will be revised to be consistent with Figure 5.4-7.
SSAR Revision:
Figure 5.4-6 is revisea to show the motor-operated valve in the IRWST discharge to the normal residual heat remosal sy stem suction line in the normally-closed position as show n in the following-a W WestinEhouse 440.129-1
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NRC REQUEST FOR ADDITIONAL INFORMATION
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-w w-w n
s.
I i w-w-vi Figure 5.4-6 Normal Residual Heat Removal System 440.129-2 W WestinEhouse
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NRC REQUEST FOrl ADDITIONAL. INFORMATION
=- m Question 440.135 Section 5 4.7.4.3 of the SSAR states that, if need arises, the RNS has the capability of transferring water from the IRWST to the refueling casity during refueling, which is performed by the spent fuel pit cooling sy stem. Is this function performed by the RNS through the direct s essel injection line?
Response
Yes. If the normal residual heat removal system is used to transfer the contents of the in-containment refueling u *-.torage tank (IRWST) to the Refueling Casity during Refueling operations, the flow path would be from the 1.
" to the normal residual heat removal system pumps and then to the refueling cavity via the direct sessel injeumn lines into the reactor sessel.
SSAR Revisiorr
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440.135-1 W Westingtlause
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NRC REQUEST FOR ADDITIONAL INFORMATION
- r..A Question 440.139 To pres nt an inadsertent closure of the RNS suction isolation takes u hile the RNS is in use (Generic Issue 49),
Section 7.6.1.1.1 of the SSAR states that their motor breakers are opened or remosed during extended RNS ope,tions following cooldown. Are opening or removal of the motor breakers done automatically or manually ?
' ill this manual operator action be included in the operating procedure?
Response
The opening of the motor breakers of the normal residual heat removal system suction isolation vahes will be performed manually. This operator action will be included in the combined license applicant operating procedures, SSAR Revision: N O N F.
4 0. m W Westinghouse 1
NRC REQUEST FOR ADDITIONAL INFORMATION iii!I li l
l Question 440.157 l
hir ref ueling operations. TS l.CO U).2 in Chapter 16 of the SS AR specifies that each valve used to isolate unborated water sources shall he secured in the closed pos,ition, whereas Section 9.3.6.4.5.1 of the SSAR states that atlounistratise controls are used to present boron dilutions by scritying the valves in the line from the deminerali/ed w ater system are closed and locked. Clarily whether technical specifications or atiministrative controls are used to ensure that these valves are closed.
Response
The statement wniained in Section 9.3.6.4.5.1 of the SS AR requiring administrative contron to be used to prevent boron dilutions by veritying the valves in the line from the deminerali/ed water system are closed and nieked is j
gmerned by the technical specitication LCO 3.9.2 in Chapter 16 of the SSAR.
l The SS AR Section 9.3.6.4.5.1 will be revised to reference the technical specilications.
1 SSAR Revision:
The tilth paragraph in Section 9.3.6.4.5.1 will be revised as follows.
For dilution events during shutdown. the somee range flux doubling signal ic used to isolate the line from the deminerali/ed water systein by closing the two salety-yrmle -related. remotely operated valves (CVSN136A, CVS-t Vl36B). The three-way pump suction control salve (CVSW115) align:. the makeup pumps to take suction troir the l
boric acid tank anti therefore stops the dilution. For refueling operations, miminivratiw%mtrols technical specilications are used to prevent boron dilutions by veritying the valves in the line from the demineralized water sy stem are closed and locket).
W Westinghouse 1
NRC REQUEST FOR ADDITIONAL INFORMATION l
Question 440.159 Sectnin 1.2.1.2 of the SSAR states that the spring loaded pressuri/er safety valves >. tat discharge to the containment atinosphere are provided for userpressure protection of the RCS, How does the design of these valves address the concern of TMI Action Itein ll.K.3.2 regarding the probability of a sinall break LOCA caused by a stuel open saf ety rehet valse?
Response
TMI Action item II.K.3.2 addresses the probability of a small-break LOCA caused by a stuck-open power-operated rebel valve and not saf ety valves. The APNU does not contain pressuri/cr power-operated relief valces. SSAR section 1.9.3. Th ce Mile Island issues, discusses this issue for the APNO.
None Q Westin, house 440.159-1 a
e NRC REQUEST FOR ADDITIONAL INFORMATION pp li!
Ouestion 440.161 Section 6.3.2.1.1 of the SSAR states that there are pimisions m the passive RilR heat exchanger to allow the operators to open the shiehled manual valves to locally sent noncondensable rases collected in the PRHR heat e xchangers. 10 CFR 50.44(c)(3Hiii) requires that high point vents for the reactor sessel head and other systems required to maintam adequate tore cooling shouhl be remotely operated f rom the control room.
Discuss conlonnance of the PRilR llX high point sents to 10 CFR 50.44teH3Hiii).
a
~
h.
W hat is the quality classification of vahes, piping, and equipment for the l'RilR llX diwharpe icturn hne to the IRWST! Jusuly this classification.
Respt,r.a a) Remote operation of the PRI1R IIX vent valves is not necessary in the AlWHi to maintain adequate coie cooling since the ADS proudes another salety-related system that can provide long tenn core cooliny. In the unlikely case of noncondensable gas accunmlation in the PRilR llN, the ADS will automatica'iy actuate on high hot leg temperature coincident with low steam pencrator level.
b) The classitication of the valves, piping and equipment for the PP5!R llX diwharpe return line to IRWST is shown on the PXS P&lD (SS AR ligure 6.3-2),
This classiliention is in conlonnance with the AlWM)
Classilica.on System (section 3.1.2 of the SSAR), in particular:
Class A is defined for the pipe untn 3/8" oritice at the top of standpipe, Class it until the second isolation valve (valve included) and Class E 1 rom the second isolation valve to the IRWST.
This classilication is consistent with the nonsalety-related lunction provided by the PRilR llX noncondensable gas vent. The required isolation harriers for the primary pressure detennine the above classilication. The 3/8" llow testrictor allows transition f rom sately class i piping to safety class 2 and two normally closed valves in series proude the transition trom safety class 2 to non nuclear safety class.
440.161-1 W WestinEhouse
NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 460.19 The responses to these questions state that there witi be no revision to the SSAR to iramperate the contents of the respons.. L\\JJ responses should be incorix> rated in appropriate SSAR sections.
Response
Previous responses to requests for additional information in the 460 review area (liftluent Treatment) included a number of responses that identified SSAR revisions. 'Ihe other responses were in the nature of clarifications or explanations that did not require SSAR Revisions. All NRC Requests for Additional Information and the Westinghouse responses will be added to the SSAR as an appendis.
SSAR Revision:
An appendix containing the NRC Requests for Additionalinformation and Westinghouse responses will be added with Revision 2 of the SSAR.
60.1S1 W Westinehouse
=
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NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 400.23 l'rovide an iter., by-item demonstration of compliance with the guidelines of itG 1.140 for the containment filtration system.
Response
The containment air filtration exhaust subsystem is designed and tested in accordance with ASMI! N509-1989 and ASMll N510-1989, as discussed in SSAR Subsection 9.4.7. SSAR Appendix I A describes how the AlWX) nuclear filtration systems comply with RG 1.140 on an item by item basis.
SSAR Revision: NONii PRA Revision: NONii i
[ W851]Dgh0US8
NRC REQUEST FOR ADDITIONAL INFORMATION A
Question 480.62 There appears to be conflicting information ~garding the closure time limits for steam generator isolation vahes:
Table 6.2.3-1 of the SSAR indicates a 5-second closure time limit, Section 6.2.3.3.H of the SSAR indicates a 10-second time limit. Clarify the discrepancy.
Response
Table 6.2.3-1 provides the correct closure time limit for the main steam isolation and bypass valves Section 6.2.3.3.11 w ill be i.crrected as indicated below. Also section 10.3.2.2.4 is corrected to indicate the 5 second closure capability.
SSAR Revision:
Resise the third paragraph under 11 in Subsection 6.2.3.3 as follows:
The main steam-line isolation vah es, main steam line isolation vah e bypass vah es main feedwater isolation s alves, steam generator blowdow n system isolation vahes, and piping are designed to prevent uncontrolled blowdown from more than one steam generator. The main steam-line isolation valves and main steam-line isolation valve bypass valves close fully within 40 5 seconds after steam line isolation is initiated. The blowdown rate is restricted by steam Cow restrictors located witain the steam generator outlet steam nonles in each blowdown path. For main steam-line breaks upstream of an isolation vahe, uncontrolled blowdown from more than one steam generator is prevented by the main steam-line isolation valves on each main steam line. The startup feed line is connected to the main feed line outside of containment.
Revise the third paragraph in Subsection 10.3.2.2.4 as follows:
l The main steam isolation vahes c!ose fully and remain fully closed within 40 5 seconds of the receipt of a manual or automatic signal. Upon receipt of the closing signal, the main steam isolation valves complete the closing cy cle d, spite loss of normally required utility services 1such as electrical power and instrument air) for actuator and/or istrumentation. On loss af actuating power the vahes fail to the closed position. Position indication and remete manual operation of the isolation s alves are provided in the control room and remote shutdown w orkstation.
Separation of redundant control, power, and indication channels is provided for the isolation sahes. The valves are desiped to prevent meidental closure on temporary loss of electrical power and instrument air. Additionally, prosisions are made for inservice inspection of the isolation valves.
com W Westinghouse
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i
7 NRC REQUEST FOR ADDITIONAL INFORMATION Question 480.72 This question pertains to Westinghouse's statement of conformance to paiagraph 6.2.3 of the Standard Review Plan,
" Secondary Containment Functional Design " that is identified on page 6-12 of Revision I to WCAP-1305 4, " AP6(M)
Compliance with SRP Acceptance Criteria."
Clarify the comment. If the secondary containment is used for post-accident dose calculations, doesn't that mean complying with 10 Cl R Part 1(W) guidelines? If so, shouldn't that require safety grade components? It was the staff's understanding that no credit was taken for secondary containment function during a DB A. Is this so?
Response
The statement in the WCAP 13054, "AP600 Compliance with SRP Acceptance Criteria" related to conformance to SRP 6.2.3 Secondary Containment Functional Des;gn" is iacomplete. The Conunents' Somn.ary of Exceptions should re2.d " AP600 does not include a safety related secondary containment; no dose mitigation effects are credited for any hohlup in the At.xiliary I!ailding in demonstrating c amr'iance with the 10 CFR 100 guidelines. A sesondary containment consisting of portions of the auxiliary building has been credited in the AP600 Probabalistic Risk Assessment Report dose analy sis discussed in Chapter 11 of the PRA." The next revision of WCAP 13054 will incorporate the above correction.
SSAR Revision: NONE I
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[ W85tingh00S8 l
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NRC REQUEST FOR ADDITIONAL INFORMATION A
(
Ouestion 490.4 Will existing empirical data for VANTAGli-5, OFA, and STD fuel types be utilized to analyze AP600 fuel assemblies? If so, justify the use of this data.
Response
I nel rod perfonnance and design esaluations for APMK) performed using the NRC approved Westinghouse fuel performance models documented in WC AP-10851-P-A (Reference 7 of Subsection 4.2.5). Many of these fuel performance models base been empirically derived based on data obtained for Westinghouse fuel operating under a wide range of conditions, including typical PWR operation and high power test rod irradiations. AP600 fuel operation is bounded by these operating conditions. It is appropriate to apply these approsed fuel performance models to AP600 design and analysis.
SSAR Revision: NONii l
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OM W Westinghouse i
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NRC REQUEST FOR ADDITIONAL INFORMATION A
Questicn 490.6 Are the /,P600 bottom and top nozzles, guide thimbles, and grid assemblies design features the same as previous Westinghouse design, or are they NRC-approved designs (VANTAGE-5, STD etc.) modified specifically for AP600? Clarify the design. If they were modified specifically for AP600, detail all differences and providejusti-fication for these differences (Section 4.2.2.2).
Response
The basis for the AP600 top and bottom nozzles, guide thimbles and grid assemblies is the VANTAGE-5H design.
To accommodate the somewhat larger guide pin located in the reactor internals core plates, the diameter of the mating top and bottom nozzle guide pin holes has been increased from 0.875 to 0.935 inch. This change provides for proper lit up of the fuel assembly in the core. This hole diameter has negligible effect on design mrgins for both top and bottom nozzles. The mechanical design of the guide thimble design for AP600 is the same as used for V ANTAGE-5H with the exception of longer length fuel assemblies. The AP600 low pressure drop grid 3 are the same as that used on the VANTAGE-5H design.
SSAR Revision: NONE l
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l "U#
W Westinghause
l 1
NRC REQUEST FOR ADDITIONAL INFORMATION
__4 l
l Ouestion 490.7 in Section 4.2.3 of the SSAR, the last paragraph states that fuel perfornumee evaluations are documented and maintained under an appropriate control sy stem. Describe how these evaluations are handled.
Response
The Westinghouse control system is documented in the NRC approved Westinghouse Quality Assurance Program Plan (Reference 2 in Subsection 17.3.1, WCAP-8370, Revision 12A).
SSAR Revision: NONE i
l OR W WestlnEtiouse
NRC REOL'EST FOR ADDITIONAL INFORMATION A
Question 490.8 Section 4.2.3.1.4 of the SSAR states that the same rod bow model used for the 14x14,15x15, and 17xl7 type cores was also used to analyze the AP600 rod bow. Is the AP600 rod bow comparable to the above stated fuel types 7 fixplain this statement both qualitatisely and quantitatively.
Response
The f uel rods to be used in the AP600 core base the same clad dimensions and spacings betv een grids as the current 17 x 17 VANTAGE 511 designs. Thus the rod bow penalties applicable to those designs are applicable to the AP600 fuel rods. The applicable penalties are given in SSAR Subsection 4.4.2.2.5.
SSAR Revision: NONE 1
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l 490.8-1 WBS'Ingh00Se
NRC REQUEST FOR ADDITIONAL INFORMATION
=
A, Question 490.9 Section 4.2.3.6.4 of the SS AR states that corrosion of the material exposed to the coolant is quite low, and that the potential for the interference with rod cluster control assembly mosement due to possible corrosion phenomena is very low. What does "quite low and very low
- mean? Clarify these terms and provide technical justification for this statement.
Response
The design limit for metal wastage from corrosion is controlled to preclude metal build-up which would be detrimental to product performance. For the AlWX) guide thimble the decrease in inner diameter is predicted to be less than 0.5 percent. During plant start-up and operation, the rod cluster control assemblics traverse the guide s
thimble path to provide a) reactivity control due to changes in power requirements and b) rapid shutdow n of the core associated with " scramming" of the rod cluster control assemblies. No reactivity control performance problenu have been experienced in operating plants due to guide thimble corrosion build-up. The rod drop times are measured prior to criticality after a removal of the reactor head (see sun eilbmee requirerrent 3.1.5.3 in Section 16.1 ).
The testing of rod dr,p times serifies that reactivity control performance is not degraded due to guide thimble corrosion build up.
SSAR Revision:
Revise the second paragraph of Subsection 4.2.3.6.4 as follou s:
Corrosion of the materials exposed to the coolant is controlled such that metal build-up which would be detrimental to product performance is precluded, istmt#4awramb, Proper control of chloride and oxygen in the t
coolant minimirca potential for the occurrence of stress corrosion. The potential for the interference with rod cluster control assembly mosement due to possible corrosion pl..:nomena is minimized, wy4+wr-This is supported by the experienee in operating plants i m reactivity control performance problems due to guide thimble corrosion build-up.
[ W85tingh0USS
NRC REQUEST FOR ADDITIONAL INFORMATION
- -
l Question 490.10 Section 4.2.3.6.4 of the SSAR states that waterlogging is not a failure mechanism associated with AP600 control rods. Provide technical justification for this statement.
Response
Waterlogging has not been observed as a failure mechanism for Westinghouse designed rod cluster control assemblies, The ab',orber rod assembly for AP600 is identical to that used in esisting Westinghouse plants.
Therefore, failure of the AP600 control rod absorber assembly due to waterlogging is not predicted to occur.
SSAR Revision: NONE 0.10-1 W8Silligt100Se
NRC REQUEST FOR ADDITIONAL INFORMATION
=--
&=;
Ouestion 491.1 Are the uncertainties associated with the calculations of the AP600 reactivity coefficients different from those used in calculations of other Westinghouse PWR designs (Section 4.3)?
Response
The uncertainties associated v,ith the calculation of reactivity coetficients for the APbOO are not different from those used in calculations of other V' stinghouse PWR designs. The AP600 fuel assembly design is neutronically similar to comentional 17x17 VANTAGE HYBRID (VANTAGE 5H) fuel assembly designs, see WCAP-10445-NP-A, Addendum 2-A, (Reference I of Subsection 4.1.1). which, in turn, evolved from a number of previous designs.
(Refer to Section 4.1 of the SSAR).
As discussed in Subsection 4.3.3.2, calculation-measurement comparisons have been made to operating reactor data measured during startup tests and during normal power operation. These comparisons include a sariety of core geometries and fuel loading patterns, and incorporate a range of fuel enrichment, burnable absorber loading, and cycle burnup.
SSAR Revision: NONE 491.1-1 W Westinghause
NRC REQUEST FOR ADDITIONAL INFORMATION Question 491.2 Section 4.3.2.2.6 of the SSAR states that normal operation of the plant assmnes compliance with certain conditions.
What are these conditions and are they the same as those associated with other Westinghouse plants. Describe and justify any differences.
Response
As stated in Section 4.3.2,2.6 of the SSAR. " Normal operation of the plant assumes compliance with the following conditions.
Control rods in a single bank move together with no individual rod insertion diffe..ng trom the bank demand position by more than the number of steps identified in the technical specifications.
Control banks are sequenced with overlapning banks.
The control banks insertion limits are not violated.
Axial pow er u._,sibution control procedures, w hich are given in terms of flux difference control and itrol bank position, are observed."
These are the same as those associated with other Westinghouse plants.
SSAR Revision: NONE 491.2-1 W westinghause
NRC REQUEST FOR ADDITIONAL INFORMATION r.
Question 491.3 Section 4.3.2.4 of the SSAR states that the shutdown margin for the AP600 core includes an allowance of 7 percent for a ialy tical uncertainties. Is this typical of all Westinghouse designs, or is it a AP600 characteristic? If it is not typical, explain the differences.
Response
The 7 percent allowances for calytical uncertainties is typical of nutn) Westinghouse designs utilizing Ag-In-Cd rod cluster control assemblies (as is the case with the AP600). The use of this allowance has been approved by the NRC (WCAP-4217, Reference 17 of Subsection 4.3.4) and has been applied to operating plants.
SSAR Revision: NONE t
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491.3-1 W Westingtiouse
NRC REQUEST FOR ADDITIONAL INFORMATION r.-....,A Question 491.4 The subject of skewed flux distribution is discussed in Section 4.3.2.4.3 of the SSAR. Is flux slewing expected to be different for the AlWK1 compared to other Westinghouse designs? Describe and explain any dif ferences.
Response
The phenomena of axial flux redistribution is not expected to be dif ferent in the APtdX) relative to other Westinghouse designs. The AlWM) maintains a consentional 12 foot active core height and operates with a power distribution control strategy w hich regulates axial powec distributions tthus xenon distributions) in a manner analogous to other Westinghouse designs.
SSAR Revision: NONi!
NRC REQUEST FOR ADDITIONAL INFORMATION A
(
Ouestion 491.5 Section 4.3.2.4.5 of the SSAR states that the rod insertion limit is set by the rod travel limit, and that a conservatisely high calculation of the inserted worth is made, exceeding the normally inserted reactivity. Describe what is entailed in this calculation.
Response
Control rod mosement is utilized in the AlWX) in lieu of baron concentrations changes for load follow capability.
To allow sufficient flexibility in power distribution control, a cot.servatise estimate of 2 000 is employed for the rod insertion allow ance. As is the case for current core designs, this conservatism is accounted for both in the determination of rod insertion limits and in the evaluation of shutJown marrin.
SSAR Revision: NONii l
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491 5-1 W Westinghouse
NRC HEOUEST FOR ADDITIONAL INFORMATION
-.. _A Question 491.6 I
lloron concentrations for various core conditions are presented in Table 4.3-2 of the SSAR for the initial cycle.
How do these concentrations vary from other Westinghouse designs (Section 4.3.2.4. Ii)?
Response
Initial core baron concentration will vary based upon a number of parc. meters including core average enrichment, I
burnable absorber content, and control rod worth. The AP600 initial cycle boron concentrations for the various core conditions as presented in Table 4.3-2 are typical of Westinghouse designs. A comparision to a 157 assembly 17x17 fir *.t core design with an initial core average enrichment at approximately 2.4 w /o U-235 is provided below.
Comparison of lloron Concentration for Various Core Conditions lloron concentrations (ppm) 157 Assembly Core Conditions AP600 17x17 W Design Zero power, Lerr = 0.99, cold, RCCAs out 1304 1319 Zero pow er, Lerr = 0.99, hot, RCCAs out 1218 1221 Zero power, L r G 0.95 cold, RCCAs in 1108 1130 eg Zero power, Lerr = 1.00, hot. RCCAs out 1121 1145 Full pow er, no senon, kerr = 1.0, hot, RCCAs out 1020 1022 I ull power, equilibrium xenon, L = 1.0, hot, RCCAs out 742 746 SSAR Revision: NONii W Westinghouse l
NRC REQUEST FOR ADDITIONAL INFORMATION
=
Question 492.1 Are the sensitivities, standard deviations and uncertainty allowances for each parameter in the revised thermal design procedures associated with the AP600, the same as those previously applied to Westinghouse designs? Provide justification (qualitatively and quantitatively) for the values of the parameters used for the AP600 design (Section 4.4).
Response
- The sensitivities used in the revised thermal design procedures analysis were calculated explicitly for the AP600 The standard deviatina and uncertainty allowances for power, temperature, pressure and flow were assumed core.
to be those used for typical Westinghouse plants. Final salues of these standard deviations and uncertainties cannot be obtained until after the actual plant is built since they are obtained from plant measurements. Variations from the assumed s alues should not cause a change to the departure from nucleate boiling ratio (DNBR) safety analysis limits, The DNHR safety analysis limits for a specific plant will be submitted by the Combined License applicant.
The standard desiations and uncertainty allowances used for bypass flow,1:N
, and THINC IV and transient code uncertainties correspond to those used for Westinghouse plants. These values are not plant specific.
SSAR Revision: NONE W WestinEhouse m
NRC REQUEST FOR ADDITIONAL INFORMATION A
Question 492.3 Section 4.4.2.2.5 of the SSAR states that a madmum rod bow penalty of less than 1.5 percent DNitR is accoun.ed for in the design safety analysis. based on an average burnup of 24,(XX) MWD'MTU. Prosidejustification for the 1.5 percent DNilR value.
Response
The fuel rods to be used in the AlbOO core have the same clad dimensions and spacings between grids as the current 17x17 V ANTAGli 511 designs. Thus the rod bow penalties applicable to those designs are applicable to the AP600 fuel rods. The applicable penalties are given in SSAR Section 4.4.2.2.5 and supported by References 20 and 21 of Subsection 4.4.7.
SSAR Revision: NONii 1
492.3-1 W westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 492.4 Section 4.4.2.2.1 of the SS AR addresses the W-3 DNil correlation. It states that, using additional information, the W-3 correlation is show n to be applicable with pressures of 300-500 psia. It is not clear w hat is meant and inferred by " additional information" Provide clarification and justification identifying the need and use of additional in formation.
Response
The pressures associated with some of the AP600 steamline break statepoints are in the range of 300 to 500 psia.
The %-3 correlation was justified to pressures down to 500 psia in Reference 7 of Subsection 4.4.7 with a 1.45 correlation limit. Just;fication of the W-3 correlation to pressures down to 200 psia was done by comparing the W-3 correlation to the data of Reference 492.4-1 and to the Stacheth (Reference 292.4-2)and EPRI(Reference 292.4-3) correlations. These three comparisons showed that the W-3 predictions are conservative w ben extended to pressi.res as low as 200 psia when a correlation limit of 1.45 is used.
References 492.4-1 " Scientific Committee of the Academy of Sciences of the U.S.S.R. 1975, 1977 Recommendations for Calculating the Boiling Heat Transfer Crisis of Water in Uniformly Heated Round Tubes" (in Russian),
Scientific Committee, Academy of Sciences. Heat and h1 ass Transfer Section, hioscow.
492.4-2 R. V. Alaebeth. " Burnout Analysis; Part 4. Application of a lecal Conditions Hypothesis to World Data f
for Uniformly Heat Round Tubes and Rectangular Channels," AEEW-R267,1963.
492.1-3 D. G. Reddy and C. F. Fighetti, " Parametric Study of CHF Data Volume 2: A Generalized Subchannel CHF Correlation for PWR and BWR Fuel Assemblies," EPRI NP-2609, January 1983.
SSAR Revision: NONE l
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l 492.4-1 W Westingtlause l
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7 NRC REQUEST FOR ADDITIONAL INFORMATION A
Question 952.78 What is the expected volume and Huid temperature of the lower plenum of the SPES-2 facility that will become stagnant at the very bottom of the pressure vessel?
Response
The response to this RAI is provided in the response to RAI 952.32.
SSAR Revision: NONE PRA Revision: NONE
~ ~ '
952384 Westinghe'Jse
NRC REQUEST FOR ADDITIONAL INFORMATION Question 952.83 What are the bend' elbow radii in the following piping systems in the SPliS-2 facility:
Downcomer-Upper Head bypass (pg. 30): 2 elbows a.
b.
Cold leg to CNIT balance lines (pg. 32): 13 elbows c.
CAIT Discharge lines (pg. 33): 10 elbow s,4 bends d.
IRWST Injection lines (pg. 34): 14 elbows e.
l'RHR HX, Supply line, Return line (pg. 35): 7 elbows f.
Accumulator injection lines (pg. 36): 8 bends or elbows
Response
The response to the RAI is provided in the response to RAl 952.29 SSAR Revision: NONI!
PRA Revision: NONti l
e s2.83.,
W westngnouse 1
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NRC REQUEST FOR ADDITIONAL INFORMATION Al Ouestion 952.85 Clarify the position relationship of the pressure vessel connections of DVI A and 11 of the SPES-2 facility with respect to the pressure sessel cold leg noule attachment locations. Is DVI A between Cold Legs A2 and ill or between Cold I.eps 132 and Al?
Response
The response to this RAI is prosided in the response to RAI 952.33.
SSAR Revision: NONii PRA Revision: NONE l
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952.85-1 3 Westinghouse l
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NRC REQUt.ST FOR ADDITIONAL INFORMATION
=4 Question 952.87 Page 8 of I)w g. (Kil891)l>92 show s a cross section of the upper plenurn and annular downcoiner in SPl.S2. 'lhe H circuinferential pressure drop fins and the 2 hot leg baf fle plates can be seen. Ilov -ver, there are 4 other structures s.how n in this crosvwetional view that correspond t the saine sessel atiinuthas ngles as the cold leg noules. What are these desiees iheir physical disnensions (including elevation in the ant. alar dow ncomer), and their purpow in the SPI K2 facility, including their relatiot.. hip to any feature or expecte i phenomena in the apt 4Krs downcomer or vessel?
Response
The response to this R AI is provided in the response to RAI 952.36.
SSAR Revision: NONii PRA Revision: NONI!
952.87 1 W Westinghouse