ML20029D256
| ML20029D256 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 04/28/1994 |
| From: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Borchardt R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NTD-NRC-94-4110, NUDOCS 9405040350 | |
| Download: ML20029D256 (28) | |
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i Westinghouse Energy Systems Electric Corporation
$3f5 samaa.oaz NTD-NRC-94-4110 DCP/NRC0044 Docket No.: STN-52-003 April 28,1994 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION: R.W.BORCHARDT
SUBJECT:
WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION ON THE AP600
Dear Mr. Borchardt:
Enclosed are three copies of the Westinghouse responses to NRC requests for additional information on the AP600 from your letters of January 26,1994, January 27,1994 and March 16,1994. In addition, a revision of a previous response is included.
A listing of the NRC requests for additional information responded to in this letter is contained in l Attachment A. Attachment B is a complete listing of the questions associated with the January 26, l 1994 and January 27,1994 letters and the corresponding letters that provided our response.
These responses are also provided as electronic files in Wordperfect 5.1 format with Mr. Hasselberg's l copy.
If you have any questions on this material, please contact Mr. Brian A. McIntyre at 412-374-4334.
1: -- l
&s ~JR,,]3.Q Nicholas J. Uparulo(Manager Nuclear Safety & Regulatory Activitics 1
/nja Enclosure cc: B. .A. McIntyre - Westinghouse F. Hasselberg - NRR uvGU g f A q ."
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I 9405040350 940429 PDR ADOCK 05200003 A PDR
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. NTD-NRC-94-4110 ATTACHMENT A AP600 RAI RESPONSES SUBMITTED APRIL 28,1994 RAI No. Issue 220.027 Potential sources of missiles in containment 220.033 : Electrical penetration assembly strength 220.037 : Containment shell prebuckling stresses 220.047 Analysis methods for seismic Cat. II structures 220.049 : Exclusion of Cat II structures for foundation anal 220.059 : Radius and thickness of dome 220.068 : Containment shell yield stress properties 230.035 : Results of 2D SSI & 3D response spectrum analyses 230.047 : Stability of containment vcessel during SSE 230.048 l Description of" design by rule" analysis 230.051 l COL commitment for reconciliation analysis 230.057 : Integrity of cont. shell-shield bldg connection 440 034R01:' CMT scaling tests 952.042 , Pressurizer Balance Line Piping Diagrams 952.043 : RELAP-5 Thermal Hydraulic Information I
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. ATTACHMENTB CROSS REFERENCE OF WESTINGHOUSE RAI RESPONSE TRANSMITTALS TO NRC LETTERS OF JANUARY 26,1994 AND JANUARY 27,1994 Question issue NRC Westinghouse No. Letter Transmittal Date 220.024 Wind-induced failure of nonsafety structures 01/26/94 04/14/94 220 025 Containment seals at transition region 01/26/94 03/264 220 026 Structuralintegrity testing of steel containment 01/26/94 03/064 220.027 Potential sources of missiles in containment 01/26,94 04/28/94 220 028 Loading effects of air baffle on containment 01/26/94 04/164 220.029 Use of (1.,0.4.0.4) method vs SRSS method 01/26/94 04/1494 220.030 Justification for factor of safety of 1.67 01/26/94 03/2494 220 031 Stress calculation by ASME criterion 01/26/94 03/064 220.032 Justifcate for factor of safety of 2.5 01/26/94 03/2494 220.033 Electrical penetration assembly strength 01/26/94 04/28/94 220 034 Nonmetalic items under SA conditions 01/26/94 03/24/94 220.035 Corrosion allowance for steel containment plates 01/26/94 03/04/94 220.036 Coritainment shell stress analysis results 01/26/94 03/24/94 220 037 Containment shell probuckling stresses 01/26/94 04/28/94 220 038 Axisymmetric rnodel vs. Sandia criterla 01/2694- 03/24/94 220 039 Strains at discontinuities vs. Sandla criteria 01/2694 03/2494 220 040 Concrete cracking effects in seismic analysis 01/2& 94 04/21/94 220 041 Soil pressure effects on embedded wall section 01/26/94 04/1494 220 042 Design criteria for severe weather phenomena 01/26,94 03/2M4 220 043 Stabihty evaluations for safety-related structure 01/26/94 03/2M4 220 044 Methodology for seismic load calculations 01/26/94 03/24/94 220 045 Subcompartment global pressure / temperature effects 01/2694 04/1494 220 046 Use of epoxy-coated reinforcing steel 01/26/94 03/24/94 220.047 Analysis methods for seismic Cat. Il structures 01/26/94 04/28/94 220.048 Capability of connection, reinforcement pattom 01/26/94 04/14/94 220.049 Exclusion of Cat il structures for foundation anal 01/2&D4 04/20/94 220.050 Factor of safety for shding & overtuming 01/26/94 03/2494 230 024 Difforence between non-Cat I & non-seismic 01/26/94 03/24/94 230 025 Non-Cat I & seismic Cat il ctarificate 01/26/94 03/24/94 230 026 GDC for seismic Cat ll 01/26/94 03/064 230 027 Frequency intenrels in response spectra 01/26/94 03/2494 230 028 Ground motion cross correlation coemcients 01/2& 94 03/2494 230 029 Basis for damping ratio 01/26/94 03/24/94 230 030 Basis for hard-rock, soft-rock damping values 01/2694 03/24/94 230 031 Shear wave velocity profile for base rock 01/26/94 03/24/94 230 032 Location of input ground motion 01/26/94 03/264 230.033 Justification for envelope of potential sites 01/26/94 03/24/94 230 034 Use of "tsme history anatysis* 01/2694 03/24/94 230 035 Results of 2D SSI & 3D response spectrum analyses 01/26/94 04/28/94 230 036 SASSI code validation pachage 01/26/94 03/2494 230 037 Cutoff frequencies of fixed base model 01/2694 04/164 230.038 Seismic Cat i structures in stick model 01/26/94 04/14/94 230.039 Live loads in modeling shield & auxiliary building 01/2694 04/1494 230 010 Modehng of steel containment shell 01/26/94 03/264 230 041 Basemat in SSI analyses 01/26/94 04/14/94 230 042 Rtructural member forces used for design 01/2694 03/24/94 230 043 hsctepancy between Sectiona 3.7.2.5 & 3.7.2.1.2 01/26/94 03/2494 230 044 / pphcation of 3 components of earthquake motion 01/26/94 03/2 6 4 230 045 A nalyses for fixed base structural model 01/2& 94 03/2494 230 048 Exclusion of addmonal accidental torsion 01/26/94 04/14/94 230 047 Stability of containment veessel during SSE 01/26/94 04/28/94 230.048 Description of " design by rule" analysis 01/26/94 04/2694 230 049 Modeling procedures 01/26/94 04/1494 952 042 Pressurizer Balance Une Piping Diagrams 01/27/94 04/28/94 952.043 RELAP 5 Thermal Hydraunc Information 01/27/94 04/28/94 Recorda ponted 55 i
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NRC REQUEST FOR ADDITIONAL INFORMATION Question 220.27 Provide the potenti;d sources of a missile or sources of high pressure resulting fnim high-energy line break between the steel containment and the operaung floor and ref ueling envity walls, between the secondary shield walls and the steel containment, and between the steel containment and the shield building (Section 1.8.2 of the SS AR).
Response
Potenhal sources ol' missiles inside the containment are discussed in SS AR Subsection 3.5.1.2. Criteria are defined to detennine if certain rotating equipment or high energy systems could result in credible missiles. When the equipment is procured and detail design infonnation is available, the equipment will be reviewed against the criteria defined in SS AR Subsection 3.5.1.2. If missiles are detennined to be credible, an evaluation will be perfonned to confinn that such missiles do not jeopardize safe shutdown.
Iligh energy piping is identified in SSAR Appendix 3E. These figures show the containment boundary.
Subcompartments are designed for the pressure and temperature effects calculated for the postulated pipe breaks as described in SS AR Subsection 3.6.1.2.1. Table 220.27-1 lists the high energy piping (greater than 1 inch nominal diameter) inside each compartment within the containment, showing the nominal site of each line. The subcompartments are identitied using the room numbers and room names given on SSAR non-proprietaiy Figures 1.24 thru 1.2-10. There is no high energy piping that can pressuri/e the annulus between the containment vessel and the shield buihling. Guard pipes are provided for the inainsteam, feedwater and steam generator blowdown containment penetrations passing through the annulus as shown in SS AR Figures 3.8.24. The CVS makeup piping is classified as high energy due to its design pressure but does not cause pressuri/ation because it is at ainbient temperat ure.
SSA.R Revision: NONE W Westinghouse 22a2 N
NRC REQUEST FOR ADDITIONAL INFORMATION l l
l TAlli.E 220.27-1 AP600 SUllCO,\lPARTMENTS ANI) POSTULATED PII'E ItUPTURES COMPARTMiiNT LINES QUALIFIED TO LBB LINES NOT QUALIFIED TO ASME Class 1 and 2 LBB ROOM NUMBERS 11201, 31" llot Leg (RCS) 3" Purilication (CVS) 11301,11401,11501 22" Cold Leg (RCS) 2" SG Blowdown (SGS)
STEAM GENERATOR 18" Surve Line (RCS)
COMPARTMENT 1 12" Fourth stage ADS (RCS) 10" Passive RHR (RCS/PXS) 16" Feed Water (SGS) 4" Pressurifer spray (RCS) 4" SG Blowdown (SGS)
ROOM NUMBERS 11202, 31" Ilot Leg (RCS) 2" SG Blowdown (SGS)
I1302, 11402. I1502 22" Cold Leg (RCS)
S'I EAM GENERATOR 12" Fourth stage ADS (RCS)
COMPARTMl!NT 2 20".12" Nonnal RilR (RCS/RNS)
" Feed Water (SGS) 8" CL to CMT (PXS) 4" SG Blowdown (SGS)
ROOM NUMBER I1205 ll" flot Leg (RCS) None REMTOR VESSEL NOZZLF "" Cold Leg (RCS)
AREA S" Direct Vessel Injection (RCS)
ROOM NUMBER I1206 S" Direct Vessel injection 2" CMT (PXS)
PXS VALVE AND (RCS/PXS) 2" Accumulator (PXS)
ACCUMULATOR ROOM A S" line from CMT (PXS) 6" line from 1RWST (PXS)
ROOM NtJMILER 11207 8" Direct Vessel Injection 2" CMT (PXS)
PXS VALVE AND (RCS/PXS) 2" Accmnulator (PXS)
ACCUMUI.ATOR ROOM B S" line from CMT (PXS) 6" line from IRWST (PXS) 220.27-2 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION N
ROOM NUMBER 11208 12".10" Nortnal RHR (RNS) None RNS VALVE ROOM ROOM NUMBER 11209 None 3" Purilication (CVS)
CVS ROOM 2" (CVS )
ROOM ff MBER 11303 6 RCS Surge Line (RCS) None PlPE ANL VALVE ROOM 10" Passive RiiR (PXS)
(BELOW PRESSURIZER) 4" Pressuriter spray (RCS) 4" SG Blowdown (SGS)
ROOM NUMBER 11403 10" Passive RiiR (PXS) None LOWER PRESSURIZER 6" Pressurizer spray line (RCS)
COMPARTMENT ROOM NUMBER 11503 14" ADS (RrS) None UPPER PRESSURIZER 10" Passive RHR (PXS)
COMPARTMENT 6" Pressurizer spray (RCS)
ROOM NUMBER 11300 32" Main Steam (SGS) 3" Purification (CVS)
MAINTENANCE FLOOR 16" Feed Water (SGS) 2" (CVS)
(LOWER COMPARTMENT) 10" Passive RilR (PXS) 4" SG Blowdown (SGS)
ROOM NUMBER 11500 32" ID Main Stearn (SGS) None OPERATING DECK 16" ID Feed Water (SGS)
(UPPER COMPARTMENT) 10" PRHR (PXS) 14". 8" 4" ADS (RCS) 6" Pressuri/er Safety (RCS)
l NRC REQUEST FOR ADDITIONAL INFORMATION I l
4 Question 220.33 I NUREG/CR-5334 reported that, during severe accident conditions, no leakage was detected from any of the three current electrical penetration assemblies (EPAs), under the following conditions (1) D. G. O'Brien EPA, 361 F, 155 psia for 10 days (2) Westinghouse EPA,400 F,75 psia for 10 days, and (3) Conax EPA,700 F,135 psia for 10 days. However, the SSAR does not address what EPAs will be used for the AP600. Provide a commitment in the SSAR that EPS penetrating containment be at least as strong as the steel containment vessel (Section 3.8.2 of the SSAR).
Response
The electrical penetration assemblies are described in SSAR Subsection 3.8.2.1.6 and are depicted in sheets 8 and 9 of Figure 3.8.2-4. The electrical penetration assemblies are procured as equipment and the details are dependent on the supplier. The assemblies will be qualified for the containment design basis event conditions as described in SSAR Appendix 3D. De assemblies will be procured to be similar to one of those tested by Sandia as reported in NUREG/CR-5334 and will have ultimate capacities consistent with those demonstrated in the Sandia tests. The ultimate capacity of the EPAs is primarily determined by the temperature. The maximum temperature of the containment vessel below the operating deck during a severe accident is reported in Appendix L of the PRA Report as 315*F. This is significantly below the capability of the assemblies tested from the three suppliers.
SSAR Revision: NONE 220.334 W-Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Question 220.37 Submit the pre-buckling stresses for the most highly stressed regions , and verify that stresses at buckling are within the elastic range (Section 3.8.2 of the SSAR).
Response
The pre-buckling stresses under design basis canditions are discussed in the response to RAI 220.36 in both meridional and circumferential directions. The most highly stressed regions away from discontinuities for buckling are the knuckle area of the top head under internal pressure, the cylinder and top head under external pressure, and the base of the cylinder under safe shutdown earthquake. The pre-buckling stresses are within the clastic range for these locations.
The bottom head is embedded in the concrete base at elevation 100 feet. This leads to high circumferential stresses at the discontinuity under thermal loading associated with the design basis accident. Buckling close to the base is evaluated against the criteria of ASMii Code Case N-284 using a BOSOR-5 model.
Detailed stress analysis results for the containment shell are available for staff review in the design calculations for I the containment vessel SSAR Revision: NONii )
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'l 220.37-1 W Westinghouse
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NRC REQUEST FOR ADDITIONAL INFORMATION F1 AP600 Question 220.47 Specily the analysis methtals and the design criteria for seismic Category II structures (Section 3.8.4 of the SSAR).
Response
The analysis meth4(Is and design criteria for seismic Category li structures are described in SS AR Subsection 3.7.2.8.
'lhis subsection of the SSAR is being revised by the response to RAI 230.54.
SSAR Revisions: None 220.47a w westingnouse I
NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 220.49 The seismic Category 11 structures, such as the turbine building, the annex buildings I :uid 11, and the solid radwaste building ere sufficiently close to the nuclear island such that their collapse could alfect the safety function of Category I structures. The structural integrity is the requirement for seismic Category 11 structures. Therefore, provide the reason why the seismic Category Il structures are excluded for the foundation analyses (Section 3.8.5 of the SSAR).
Response
The information requested will be provided in May concurrent with the responses to RAls 230.54, 230.66, 230.68, and 230.73 which also relate to seismic Category 11 structures.
I 220.49-1 W-Westinghouse 1 i
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NRC REQUEST FOR ADDITIONAL INFORMATION Question 220.59 l'rovide the radius and the thickness of the knuckle region and the dome in Section 3.8.2 of the SSAR.
Response
Dimensions of the containtnent vessel are given on sheet 1 of SSAR Iigure 3.8.2-1. The head is ellipsoidal with a major diarneter of 130 feet and a height of 37 feet 7.5 inches. The thickness is 1.625 inches.
SSAR Revision: NONil l
l 1
220.59-1
[ W85tlngh00Se
NRC REQUEST FOR ADDITIONAL INFORMATION i Question 220.68 With regard to the materials to be used for the containment shell, a stress-strain curve for SA-537 Class 2 material was presented that was reported to be obtained from Japanese data. The yield stress was shown to be 81.3 ksi.
However, Table 2 of the ASME specification for SA-537 reports a minimum yield strength of 60 ksi for the same material. Clarify the correct properties to be used for design
Response
As described in SSAR Subsection 3.8.2.4.2.6, the containment vessel is designed using SA537, Class 2 mterial.
The design is based on ASME speci6ed properties, which at ambient temperature are 60 ksi yield stress a.id 80 ksi ultimate stress.
A typical stress strain curve is also described in SSAR Subsection 3.8.2.4.2.6. This had a yield stress of 81.3 ksi.
This value was used in calculating the best estimate containment pressure for general membrane yield of the cylinder which is assumed to correspond to the loss of containment function.
SSAR Revision: NONE 220.68-1 W Westinghouse
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j NRC REQUEST FOR ADDITIONAL INFORMATION Question 230.35 The following request for additional information pertains to Section 3.7.2.1.1 of the SSAR:
- a. Provide the detailed comparison of the results obtained from the 2D SSI analyses and the 3D response spectrum analyses for the hard. rock site condition.
- b. As described in Section 3.7.2. l.1, the structural member forces and moments are obtained from the response spectrum analysis of the finite element model for the hard-rock site, and from the SSI analysis of the stick model for the soil sites. Provide a comparison of responses from the response spectrum analyses of a stick model and a finite element model at rock site.
- c. From the staff's review of Section 3.7.2.1.1 and Table 2A.17, the staff determined that the hard-rock site condition (RI) is not the governing case for the steel containment shell. Describe how the steel containment shell was analyzed for the rock site condition.
- d. Provide the rationale for excluding the SB roof in the finite element model, as shown in Figure 3.7.2-1.
- e. From the staff's review of Tables 3.7.2-1 through 3.7.2-4 of the SSAR, the staff determined that the AP600 nuclear island structures (except the steel containment shell) are very rigid. Some predominant frequencies are much higher than 33 Hz. Provide justification for the statement "since the shear wave velocity for the hard r >ck site is in excess of 8000 feet per second, the soil- structure interaction effect is negligible " This statement has also been made in Sections 3.7.2.1.2 and 3.7.2.4.
Response
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- a. Maximum member forces for the hard rock (RI) case of the 2D SSI analysis are given in Table 2A-17.
Maximum member forces for the hard rock analpes of the 3D stick model using the computer program BSAP are given in Table 3.7.2-11 (sheet 1). Floor response spectra for the R1 case of the 2D analyses are given in Figures 2A-29, 2 A--30 and 2 A-31. Floor response spectra for the USAP hard rock analyses of the 3D stick model are given in Figures 3.7.2-29,3.7.2-30 and 3.7.2-31. The 3D stick model was developed from the finite element model and the frequencies and modal participation of the 3D stick model and finite element model are consistent. The 2D SSI analysis was performed to establish the design soil profiles for the AP600 plant. 'I'he l 3D response spectrum analysis reported in Revision I of SSAR Subsection 3.7.2 for the hard rock site condition was performed to obtain in-plane member forces in the individual elements of the finite element model. There were slight differences in the plant configuration considered in the 2D SSI model and the 3D models. More detailed comparison of results from the two analyses is not meaningful.
- b. lhe response spectmm analysis described in Subsection 3.7.2.1.1 was performed only for the hard rock site and used the three-dimensional finite element models. It generated forces and moments in the va;ious elements such as individual walls and slabs. The member forces and moments obtained in the time history three-dimensional analyses of the lumped-mass stick models (described in the last paragraph of Subsection 3.7.2.1.1) 230.3s-,
W wesueuse
NRC REQUEST FOR ADDITIONAL INFORMATION nn E F tii ,
1
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\
l are typically the total shear force, axial force, and tnoment at a given elevation in the structure. A direct comparison it. not available.
- c. Table 3.7.2-12 shows the maximum member forces in the containment vessel stick model for the three design soil conditions (hard rock, soft rock and soft-to-medium stiff soil). These results show that the hard rock case gives the maximum forces. Table 3.7.2-6 shows the maximum absolute accelerations for the same soil conditions. Le hard rock case results in the highest accelerations of the vessel, except in the node representing the colar crane where the acceleration in the east-west direction is 6% higher for the soft rock case than for the hard rock case. This is considered in design of the crane girder which uses the crane wheel loads from the polar crane design analyses. These design analyses will be reconciled by the Combined License applicant once the final design of the crane is established.
The steel containment vessel is analyzed using the shell of revolution model for the equivalent static accelerations from the SSI seismic analyses reported in SSAR Table 3.7.2-6.
He analyses of Appendix 2A are intended to select the appropriate soils cases for the 3D analyses reported in SSAR Section 3.7.2. They are not used to define the governing case for the containment vessel design. Table 2A.17 shows the seismic member forces for the containment vessel for these parametric soils analyses. This ,
data is for a configuration in which the containment vessel was supported up to elevation 82'-6" As reported in Table 2A.15 this model had a fundamental frequency of 2.14 Hz in the east-west direction. Based on review i of these results a design change was incorporated to raise concrete around the vessel to elevation 100'. This increased the fundamental frequency of the containment vessel to 7.61 Hz (see SSAR Figure 3.7.2-10). This model is included in the analyses of SSAR Section 3.7.2. The analyses of Appendix 2A are appropriate for )
the selection of soil condi9 ens because the mass of the containment vessel is small compared with that of the l rest of the nuclear island.
- d. A lumped-mass stick model of the shield building roof structure was constructed and coupled with the finite element model and the stick model on the coupled auxiliary and shield buildings, The stick model of the shield i building roof structure was included in all seismic analyses performed. The lumped-mass stick model of the shield building roof was not shown in Figure 3.7.2-1 to maintain visual clarity of the finite element model.
- e. For the hard rock site, a fixed-base analysis was performed based on the acceptance criteria specified in Revision 2 of SRP 3.7.2. "For structures supported on rock or rock-like material, a fixed base assumption is acceptable. Such materials are defined by a shear wave velocity of 3500 feet per second or greater at a shear strain of 10-3 percent or smaller ...etc." Furthermore, as noted in Section 3.7.2.2, the total cumulative mass of the nuclear island participating in the seismic response, up to the frequency limit of 34 Hz, constitute 90, j 40 and 83 percent of the total mass, excluding the building mass within the embedded portion. The '
predominant frequencies of the coupled auxiliary / shield buildings and the steel containment vessel are below 34 Hz. The relatively rigid containment internal structures, coupled to the other flexible structures on a common basemat, are expected to have negligible effect in the over all soil-structure interaction responses of the nuclear island. Herefore, for the hard rock site, only a fixed-base analysis is required.
SSAR Revision: NONE 230.35-2 W- Westinithouse
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NRC REQUEST FOR ADDITIONA1. INFORMATION
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Question 230.47 Section 3.8.2.1.2 of the SSAR states that the vertical and lateral loads on the containment vessel and internal structures are transferred to the basemat below the sessel by friction and bearing. This statement implies that there are no shear studs or anchors between the internal structures, steel containment vessel and reinforced concrete basemat. Provide an analysis to demonstrate the dynamic stability of the containment vessel during an SSE event or a seismic margins earthquake.
Response
There are no shear studs or anchors between the internal structures, steel containment vessel and reinforced concrete basemat. The dynamic stability of the containment vessel was evaluated for a SSE event using n conservative friction !
coefficient of 0.4 at the concrete / steel interface. He factors of safety computed are equal to 2.5 and 3.0 against overturning and sliding, respectively.
He evaluation is graphically presented in Figures 230.47-1 and 230.47-2 using the following input data:
- The total dead v eight of the steel containment vessel (SCV), he containment internal structures (CIS), and major equipmen<,
I W= 61,266 Kips a The peak SSE response forces and moments of the SCV and the CIS for the three design soil profiles are ,
enveloped. The enveloped SSE response forces and moments of the SCV and the CIS are assumed to occur I simultaneously and combined at Elevation 66'-6" Using the (1.0, 0.4, 0.4) method, the combined SSE ,
response forces and moments used in the evaluation are i
F, - 9,070 Kips F,, - 23,045 Kips M- 1,506,770 K-It The dynamic stability analysis for the safe shutdown earthquake shows a large safety factor of 2.5 against overturning. Based on an analysis similar to that shown in Figure 230.47-1, incipient overturning would be predicted j at a ground input level of 0.79g for an earthquake having a similar response spectrum shape as the safe shutdown j earthquake. This would not represent failure since small lift-off w ould not ,a n nt safe shutdown. Dynamic stability is therefore assured for the seismic margins review level earthquake. l l
SSAR Revision: NONE 230.47-1 W-Westlngh00Se 4 i
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NRC REQUEST FOR ADDITIONAL INFORMATION Y
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eo e-M1 H=S F 1,506,77p 23,045 65.4 ft 3
Assumed Overtuming about Point "A", the low point of the minimum bearing area required to support the dead weight + SSE ioads at the concrete cradle.
Overtuming Moment = ' nF + 41.1 + F,,
- 60.8 - 1,498,606 K-ft Resisting Moment - W + 60.8 4 R3 + 3.9 - 3,812,417 K-ft
' Factor of' Safety = Resisting Moment , g,5 _, 3.3 Overtuming Moment
. figure 230.47-1 Factor of Safety To Resist Overturning .
230.47-2 W westinghouse-
. 1 NRC REQUEST FOR ADDITIONAL INFORMATION Acouned Sliding Circle - -
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, g$,4 f, Fs 23,045 4 = Tan ~'
= E'US = 23.82
- W - F, S2,196 0 - Sin .i (96.45 - II) x Sin (180 - 4) 7,47 96.45 Reactions, nonnal and tangential, to asstuned sliding circle:
R, = F, Sin 0 = 7,422 Kips , R, =
F, Cos 0 = 56,572 Kips
( R,xp ) $6,$72 x 0,4 Factor of Sqfety = =
- 3.0 :e 1.1 R, 7,422 Figure 230.47-2 Factor of Safety To Resist Sliding W westingnotse
NRC REQllEST FOR ADDITIONAL INFORMATION r=
Question 230.48 Provide a detailed description regarding the " design by rule" analysis inethod in the SSAR and discuss what activities are underway for adoption of this method by a consensus code or standard (Section 3.7.3.1 of the SSAR).
Response
The " design by rule" method for small bore piping is based on EPRI Report NP6628 as described in SSAR, Revision I, subsection 3.7.3.8.2.2.
SSAR Revision: NONE l
1 l
i OA84 3 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION mr Ouestion 230.51 The seismic design bases for the AP600 standard design are essentially defined by a safe-shutdown earthquake (SSE) with peak grounJ acceleration of 0.3g and the soil profiles characterized in Section 2 of the SSAR (assuming no liquefaction and fault displacement at the plant site). If these generic design bases are not satisfied, design certification will no longer hold, and site-specific analyses and evaluations must be performed in accordance with the SRP Provide a COL commitment, in the SSAR, to perform this reconciliation analysis.
Response
Chapter 2 of the SSAR defines the site-related parameters for which the AP600 plant is designed. These parameters envelope most potential sites in the United States. This chapter discusses how the specific interfaces are to be used in the AP600 design. The Combined License applicant is responsible to demonstrate that the selected site meets the interface. Section 2.5 provides seismology criteria by which acceptability may be demonstrated.
l l For cases where a site characteristic exceeds the envelope parameter, it is the responsibility of the Combined i
License applicant referencing the AP600 to demonstrate that the site characteristic does not exceed the capability of the design. Thus, it is not necessary or appropriate to include in the design certification of the AP600, requirements and commitments for applicants with sites that do not meet the site characteristics for the standard design.
SSAR Revision: NONIi l
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230.51-1 3 Westinghouse 1
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NRC REQUEST FOR ADDITIONAL INFORMATION Question 230.57 One of the drawings displayed shows a physical connection between the containment shell and the shield building near the upper spring line. If the function of the connection is important, its integrity should be evaluated when the connection is subject to relative displacement (between the containment shell and the shield building) during a seismic event.
Response
The shield building, containment vessel and air baffle are shown on the General Arrangement sections in SSAR Figures 1.2-12 and 1.2-13. These Ogures show the containment air baffle and the pipe strut attaching the air baffle to the containment vessel. They also show the flexible seal between the air baffle attached to the containment vessel and the portion of the air baffle attached to the shield building. The only physical connection between the shield building roof and its anached structures, and the containment vessel and its attached structures, is the flexible seal.
The upper air baffle is attached to the shield building roof. The lower portion is attached to the steel containment.
The flexible seal, at elevation 236', accommodates the differential deflections of the containment vessel and shield building under scismic, design basis and severe accident loads. The containment air baffle and its effect on the containment vessel are described in SSAR Subsection 3.8.4.1.3.
SSAR Revision: NON!!
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1 NRC REQUEST FOR ADDITIONAL INFORMATION Ii!. ~!!!'
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Question 440.34 in its January 19, 1993 response to a question on the core makeup tank (CMT) tests dated July 21, 1992, Westinghouse states that "[t]here will be no formal scaling report for the CMT tests. Since the CMT test is a separate effects test, ..the boundary conditions for the test can be separately controlled., .[thus] there is no need for a detailed scaling report." While it is true that some conditions can be more closely controlled in a separate effects test environment than in a systems test environment, once the test starts, the conditions that evolve, such as natural cenvective flows and temperature distributions, are governed by the physical processes occurring during the test itself, including heat tnmsfer to and from the CMT and depressurization of the test k>op (simulating ADS .
I actuation). If the geometry of the test article is substantially dif ferent from the prototypic component, the thermal-hydraulic behavior of the two could be different. This is the case with the CMT test. The component in the plant has an aspect ratio (height to diameter) of about 1.7, whereas the test article has an aspect ratio of about 5. Muhi-dimensional behavior in the actual CMT, including stratification internal recirculation, and energy transport, may not be adequately represented in the test article, which looks much more one-dimensional. 'This behavior may have a substantial impact on the response of the CMT during an accident. Therefore, provide a detailed scaline analysis showing that the thermal-hydraulic phenomenology observed in the CMT test can be directly related to that expected in the plant component during the range of events where the CMT is expected to be in operation.
Response (Revision 1h l WestinghoineTopical Report, WCAP 13963, " Scaling Logic for the Core MakeupTank Test," Revision 0, provides the reque.sted scaling analysis for the Core Makeup Tank test facility. The report was provided to the NRC via Westinghouse letter NTD NRC-94-4068, dated February 22,1994. I SSAR Revision: None l
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NRC REQUEST FOR ADDITIONAL INFORMATION fit ~ jj p: ...
Question 952.42 The infonnation provided for the pressure balance line (PBL) f roin the pre 3surizer to CMT is incomplete for the r.hort angled run upstream of the pressurie.er. Provide piping diagrams in a plan view from the reactor sessel center line to the west, and show the PBL piping from the reactor vessel center line to the pressurizer.
Response
Subsequent to the receipt of this RAI, an AP600 design change was implemented which removed the pressure balance lirx from the pressuri/er to the Chfr. Therefore, the requested piping diagrams are no longer relevant to the AP600 design. This change has been discussed with NRC staff arkt will be documented in the next SSAR revision. A design change report documenting the change arxi impacts on plant safety analyses will be submitted by June 30,1994.
SSAR Revision: NONE PRA Revision: NONE i
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d NRC REQUEST FOR ADDITIONAL INFORMATION Question 952.43 Provide the following information:
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- a. Ilydraulie diameter at the tube support plates.
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- b. Ilydraulie di;uneter at the downcomer annulus obstructions.
- c. Steam drier metal volume.
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- d. Inner diameter of the feedwater J nor/les. j
- e. Feedwater line piping diagrams.
j f. Ste:un line piping diagrams.
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- g. Feedwater level control systein information.
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- h. Turbine stop valve area.
f l i. Post-trip T-avg trending information (tmhine bypass valve control).
1 In the interim plant deck developed by the staff's contractor, parameters that were developed from knowledge of the ste.un penerators typical of existing PWRs, amt from knowledge of the changes that will be incorporated in the APNX) design were used. Ahhough the NRR interim deck should provide a close approximation of the AlWX) design response, no quality assurance for the deck that is traceaNe to a primary source (i.e., WEC engineering I drawings or other documentation of the APGM) specific information) can be established without the requested infor mation.
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- j. Normal RilR piping diagrams and pump curves.
- k. CVCS makeup / letdown systent information.
- l. Pumps curves,if dif ferent from those in the Westinghouse COBRA-TRAC workbook previously transinitted to INEL.
- m. lYessurifer inlet not/le retaining basket number of holes (known hole diameter).
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NRC REQUEST FOR ADDITIONAL INFORMATION
- n. PRHR drawing illustrating the conHguration :uid dimensions of the baille surrounding each tube hundle,
- o. Plot of void versus reactivity for the core.
Response
- a. Ilydraulie diameter at the tube support plates The hydraulic di:uneter at the tube support plaies is 0.189 inches.
- h. Il draulie 3 diameter at the donneomer annulus obstructions The hydraulic diameter of each downcomer annulus obstruction is identified below:
lleicht above tube sheet Hydraulic Diameg 353.82 in 3.36 in 312,56 in 3.97 in.
292.50 in 5.44 in.
273.56 in 4.39 in.
234.56 in 4.39 in.
195.56 in 439 in.
156.56 in 4.39 in.
I 17.56 in 4.39in.
78.56 in 4.39 in.
39.56 in 4.71in.
20.(W) in 5.19 in.
- c. Steam drier metal solume The steam drier metal solume is approximately 90jNM) cubic inches with a metal mass of approximately 25500 lbs.
- d. Inner diameter of the feedwater .1-noules The AP600 ste un generator uses spray nonles not 1-noules. Each of the 34 spray nonles, located on top of the feedwater ring, has 130 holes of 0.25 inch diameter. A sketch of the spray noules was provided via Westinghouse letter NTD-NRC-94-4108 1
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k NRC REQUEST FOR ADDITIONAL INFORMATION
- c. lleedwater line piping diagrams R
Feedwater line piping sketches were provided as an enclosure to Westinghouse letter NTD-NRC-94-4108, dated 4/29/94. The piping sketches represent preliminary routings. The details are subject to change as a result of ongoing design activities,
- f. Steam line piping diagrams Main steam line piping t"tches were provided as an enclosure to Westinghouse letter NTD-NRC-94-4108, dated 4/29/94. The p9ing sketches represent preliminary routings. The details are subject to change as a resuk of ongoing design activities,
- g. Feedwater lesel control system information The requested infonnation can he found in SSAR Section 7.7.1.8. The logic is shown on Figure 7.2-1, Sheets 25 and 26.
The steam generator level control logic is sunilar to the Advanced Digital Feedwater Control System
( ADFCS) which has been implemented at several Westinghouse nuclear plants, including Catawba, Diablo Canyon, and Praitie Island.
- h. Turbine stop satse area The ApN10 has four turbine throttle tstop) valves. Each throttle valve has a nominal throat area of 3S0 square inches.
- i. l'ost-trip T-asg trending information (turbine bypass salse control)
The requested information can be found in SSAR Section 7.7.1.9 The logie is shown on Figure 7.2-1, Sheet 22.
Initially following the reactor trip, the Plant Trip mode of steam dump control will be used. This control mode is described in detail in SSAR Section 7.7.1.9.2. This mode of control is no dif ferent than that at operating Westinghouse nuclear plants.
Atter the plant conditions stabili/c at no-load following a scactor trip, the operator is instructed to switch control modes to Pressure control. This mode is described in SS AR Section 7.7.1.93. For holding conditions at no-load, this mmle is no dif ferent than that used on present Westinghouse nuclear plants. For plant cooldown, the operator will input a desired plant coe'down rate; this will be converted into a pressure setpoint for use by the control system (see SSAR section 7.7.1.93).
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NRC REQUEST FOR ADDITIONAL INFORMATION E!
Other J. Normal Rillt piping diagrams and pump cunes The AP600 Nonnal Residual lleat Removal System (RNS) system distribution and pump performance curve were provided in Westinghouse letter NTD-NRC-94-4108, dated 4/29/94. This letter provides calculated flow path resistances throughout the RNS rather than piping diagrams. The plant orifices will be siicd to obtain the desired overall system resistance, and thus How perfonnance, once the piping layouts are finalized.
- k. CVCS makeup / letdown system information The chemical and volume control system operates to provide reactor coolant system purification, makeup and letdown. At a constant power level, the CVS purification loop operates as a closed h>op around the reactor coolant pumps. The CVS makeup pumps and the letdown lines to the liquid radwaste system are not nonnally operating. The following paragraphs outline the makeup and letdown functions of the CVS.
Refer to SS AR subsection 9.3.6 for additional infonnation on the CVS.
On a low-pressuriier level signal (relative to the prognu'uned level) one of the chemical and volume control system makeup pumps starts automatically to provide makeup at a controlled rate of 100 ppm. The chemical and vohune control system makeup is also controlled to within a pressurifer level band following receipt of a core makeup tank actuation signal. One makeup pump is started when level reaches the low end of the hand and is stopped when level reaches the high end of the pressurifer level hand. This prevents pressurizer overfill or pressurifer safety valve lift on a best estimate basis. It also provides for reactor coolant system makeup which reduces the chance that the core makeup tanks will drain to the automatie depressurit.ation system setpoint.
The chemical and volume control system letdown, when required is taken out of the purification h>op at a point downstream of the reactor coohmt filters. On a high pressuriter level the letdown orifice isolation valves automatically open to divert flow to the liquid radwaste system. These valves automatically close on a containment isolation signal, high liquid radwaste system degasifier level, or a low pressuriter level signal.
- l. Pump cunes, if different from those in the Westinghouse COllRA-TRAC workbook previously transmitted to INEl, The AP600 RCp ump p curves were transmitted to the NRC via Westinghouse letter ET-NRC-93-4013, dated 11/12/93,
- m. Pressuriier inlet noule retaining basket-number of holes (known hole diameter).
The pressuriter inlet nonle retaining basket has 2520 holes of 0.375 inch diameter.
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NRC REQUEST FOR ADDITIONAL INFORMATION k
- n. Pitillt drawing illustrating the configuration und dimensions of the haftle surrounding each tube bundle The drawing was provided as an attachment with letter NTD-NRC-94-4108 dated 4/29/94,
- o. Plot of sold sersus reactisity for the core The following specifies the reactivity versus density curve which is applicable to Westinghouse cores, including AP600:
dbk,,,yp,(p dp 2
where 6k is the reactivity insertion and p is the spatial moderator density, gm/cc The shape of this curve (defining constants B and C) is detennined by specifying the changes in d5k/dp from 0.X to 0.4 gmhc and f rom 0.8 to 0.6 gm/ec as 0.37 and (UN respectively. The vertical position of the cune (constant A) is zero.
1 SSAR Revision: I i
- k. The fourth paragraph of section 9.3.6.3.1 (Chemical and Volume Control System Makeup linnps) will be l revised as follows.
l One makeup pump is started on a core makeup lank actuation signal. The makeup pmniwanulartedust wfety-injection +ignal, The makeup The+ pumps provide an additional injection source to contribute to the overall reliability of the makeup function during accident conditions.
The second paragraph of section 9.3.6.4.5 (Accident Operation) will he revised as follows.
One The chemical and vohune control system makeup pumpuero is initiated upon receipt of a wifety injection signal core inakeup tank actuation signal. Although these pumps do not provide a safety function, they are available to provide reactor coolant system makeup and pressurieer auxiliary spray as an additional means to improve reliability of the makeup f unction during accident conditions.
The eighth bulleted paragraph of section 9.3.6.7 (Instrumentation Requirements) will be revised as follows, Afalerip pump control - The makeup pumps controls are hicated in the main control room. On a low-pressuriier level signal (relative to the prognunmed level) one of the chemical and volume control system makeup pumps start automatically to provide makeup. The operating pump automatically stops when the pressuri/cr level increases to the correct vahie. During reactor coolant system boron changes (fuel depletion.
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NRC REQUEST FOR ADDITIONAL INFORMATION k
startups. shutdowns, and ref ueling), the o[vrator will st;ut one of the makeup pumps atter selecting the desired amount of loric acid.
One chemical and volume control systern makeup pump will start upon receipt of a core makeup tank actuation signal. The chemieni and volume control system makeup is also controlled to within a pressuriter level band following receipt cf a core makeup tank actuation signal. One makeup pump is started when level reaches the low end of the band and is stopped when level reaches the high end of the pressurizer level ha!Kl. The stop signal prevents pressurifer overfill or pressurizer safety valvo lift on a best estimate basis.
The start of the makeup pump also provides for reactor coolant system makeup which reduces the chance that the core snakeup tanks drain to the automatic depressurir.ation system setpoint.
The operators can start the second makeup pump in case the core makeup tank drain down is approaching the automatic depressurization system setpoint. The pressurizer level control setpoints at zero power (without a core makeup tank. actuation) are greater than the post core makeup tank level stop (approximately 23 percent of the Span).
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