View source for NRC-94-4236, Forwards Responses to NRC RAIs 940419,27,29,0511,26,0601 & 0608 RAIs on AP600.Revs of Responses Previously Submitted Also Provided,Completing Response to NRC

{{Adams
| number = ML20071J440
| issue date = 07/25/1994
| title = Forwards Responses to NRC RAIs 940419,27,29,0511,26,0601 & 0608 RAIs on AP600.Revs of Responses Previously Submitted Also Provided,Completing Response to NRC
| author name = Liparulo N
| author affiliation = WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
| addressee name = Borchardt R
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05200003
| license number = 
| contact person = 
| document report number = NTD-NRC-94-4236, NUDOCS 9407280162
| title reference date = 04-07-1994
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 54
}}

=Text=
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W Westinghouse               Energy Systems                                      $355,, ,          g3g333 Electric Corporation NTD-NRC-94-4236 DCP/NRC0162 Docket No.: STN-52-003 July 25,1994 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 AP(XX)

==Dear Mr. Borchardt:==

Enclosed are three copics of the Westinghouse responses to NRC requests for additional information on the AP(XX) from your letters of April 19,1994, April 27,1994, April 29,1994, May 11,1994, May 26,1994, June 1,1994 and June 8,1994. In addition, revisions of responses previously submitted is provided. This completer the response to the {{letter dated|date=April 7, 1994|text=letter dated April 7,1994}}.
A listing of the NRC 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 format with Mr. Kenyon's copy.
If you have any questions on this material, please contact Mr. Brian A. McIntyre at 412-374-4334.
                                            /
hi                b Nicholas J. Liparulo, Manager fo --
Nuclear Safety Regulatory And Licensing Activities
           /nja Enclosure cc:     B. A. McIntyre - Westinghouse T. Kenyon - NRR 4: C 4as        L, 8
I
            "^                                                                                       I 9407280162 940725 PDR     ADOCK 05200003                                                                       !
A                      PDR

i i    t NTD-NRC-94-4236                                               i ATTACHMENT A                                                 !
AP600 RAI RESPONSES SUBMITTED JULY 25,1994                                            !
P RAI No.                                  Issue                                                  [
210.048          :  SSAR section 3.7.3.5 210.063             Method of combination of dynamic responses i
210.064          :  Use of elastic-plastic method of analysis i
210.068          :  Stress criteria for active compor.ent supports 210.%9           :  Section 3.9.3.4.3, snubber operability                                      ;
210.088          :  Seismic qualification report
{
440.00lR01:         SPES test, check valve testing & PRHR testing 440.011R01i         ADS testing 440.052          :  CMT Scaling Report 440.133          i  RNS pump suction                                                            .
440.171          :  HF analysis inclusion of high point vent operation 440.245          i  Battery bank inconsistency between SS AR & PRA                              l 480.049          i  Provisions for Type C testing, Table 6.2.3-1                                ,
480.050             Type C testing of service air 480.052          :  SSAR Table 6.2.3-1 & Figure 9.2.4-1 480.056          :  Relief valves as containment isolation barriers 480.060          ;  manual vs remote manual                                                     f 480.066          :  Margin between max calculated & design cont press                           !
480.068          :  Postulated break size for subcompartment analyses                           !
480.069          :  Use ofTMD code for M&E releases 492.005          :  Fixed incore detector monitor 952.082          l  SPES-2 pipe schedule changes
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1 NRC REQUEST FOR ADDITIONAL INFORMATION                                                                                                                                                                                                                           1 i
Question 210.48 The following requests are relative to Section 3.7,3.5 of the SSAR, "liquivalent Statie imad Method of Analy sis:"
l           a. The second paragraph in this section states that single degree of freedom subsystems are designed for l                accelerations associated with their natural frequency.                                                                                                                                                  The staffs position as stated in l
Section 3 9.2.ll.2.a(2) of the SRP is that for equipment that can be modeled adequately as a one-degree-of freedom system, only the use of a static load equivalent to the peak of the floor response spectra is acceptable. Either revise this paragraph to be consistent with the staff position, or provide the basis for I                the use of accelerations esociated with the natural frequency.
: b. The third paragraph in this section states that, for multi-degree- of-freedom systems, in lieu of using the peak acceleration value, the actual frequency may be calculated and the corresponding acceleration value may be used. It is not clear whether or not the 1.5 factor is also included in this cortesponding acceleration value. Revise this paragraph to provide a clarification of this alternative.
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: c. The fifth paragraph in this section states that the equivalent static load method of analysis can also be used I                for small-bore piping. The staff's position as stated in Section 3.9.2.ll.2.a(2) of the SRP is that an l                equivalent static load method is acceptable if justification is provided that the system can be realistically i                represented by a simple model and the method produces conservative results in terms of responses.
l                l'orthermore Section 3.9.2.ll.2.a (2)of the SRP states that the design and the associated simplified analysis i
account for the relative motion between support points and a factor of 1.5 is applied to the peak                                                                                                                                                       )'
l                acceleration of the floor response spectrum. Alternatisely, the use of a static load equivalent to the peak of the floor response spectra is acceptable for piping supported at only two points. Revise this paragraph to be consistent with the staff position. or proside the basis for the use of the equivalent static load method                                                                                                                                         j of analysis for small-bore piping.                                                                                                                                                                                                                      {

===Response===
I
: a. The static equivalent load method is applied to equipment that can be adequately modeled as a one-degree -of-                                                                                                                                               l freedom sy stem.
i
: b. When the equipment frequency is calculated the f.pectral acceleration at this frequency is used in the equivalent static load method. The factor of I             not applied to this acceleration. This is documented in Reference 210.401,
: c. The amplification factor of 1.5 w ae used for piping analy sis with the equivalent static load method, except that a factor of 1.0 is used when there is an axial support. The relative motion between supports is also considered a ben significaat.                                                                                                                                                                                                                                I I
f 210.48-1 W westinghouse                                                                                                                                                                                                                                                 l l
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l NRC REQUEST FOR ADDITIONAL INFORMATION                      i l
t         ..-

==References:==

210.48-l       WCAP-WB. "Justi6 cation Of The Westinghouse Equivalent Static Analy sis Method For Seismic Quali6 cation Of Nuclear Power Plant Auxiliary Mechanical Equipment." August,1980.
SSAR Revision:
3.7.3.5        Equivalent Static Load Method of Analysis The equivalent static load method involves equivalent horizontal and vertical static forces applied at the center of gravity of various masses. The equivalent force at a mass location is computed as the product of the mass and the seismic acceleration value applicable to that mass location.
The magnitude of the seismic acceleration is established based on the dy namic response characteristics of the subsy stem. Subsystems that can be characterized as a single degree of freedom system are designed for accelerations associated with their natural frequency. Seismic acceleration values-w.eMe+4*gn =f : ::h! Jeg**e g,,,4,,                 .e1              .t y , _
                     ,,y ,          ;i                             _r 9        p e g r.7 ,, s.e           _. , are the peak acceleration values from the applicable Goor response spectra multiplied by a factor of 1.5, unless a lower factor is justi6ed.
In lieu of using the peak acceleration salue, the actual frequency may be calculated and the corresponding acceleration value may be used without ampli6 cation. In this case, the calculated frequency must be higher than that frequency related to the peak acceleration. Otherwise, the peak acceleration value is used in design. For subs) stems and components hasing fundamental frequencies of 33 hertz or greater, the zero period acceleration is taken as the scism;c accel: ration value. This is documented in Reference 23.
The equivalent static load method of analysis may be used for design of platforms, electrical cable trays and suppor%. conduits ano supports, HVAC ducts and supports. instrumentation tubing and supports, piping systems, and other substructures. This analysis is based on single span models.
The equivalent static load method of analysis can also be used for design of piping systems, and instrumentation tubing and supports. with signi6 cant responses at several vibrational frequencies. In this case, a static load factor of 1.5 is applied to the peak accelerations of the applicable floor response spectra. For piping rtms with axial supports the acceleration salue of the mass of piping in its axial direction may be limited to 1.0 times its calculated spectral acceleration value. The spectral acceleration value is based on the frequency of the piping sy stem along the axial direction. This frequency is determined from the piping mass and the axial support stiffness value. Using this method for piping systems is limited to small-bore (two inches or less in nominal pipe size) piping. The relative motion between support points is also considered when significant.
Add Reference to Subsection 3.7.5 as follows:
: 23. WCAP-9903, " Justification Of The Westinghouse Equivalent Static Analysis Method For Seismic Qualification Of Nuclear Power Plant Auxiliary Mechanical Equipment,* August,1980.
210.48 2 W
Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Question 210.63 In Tables 3.9.3 5, 3.9.3-6, 3.9.3-7, and 3.9.3-8 of th, SS AR, add a note to state that the method of combination of dy namic responses to loads is in accordance with the recommendations in NUREG-0484, " Methodology for Combining Dynamic Responses," Revision I, dated May 1980, in addition, explain how Note 6 in Table 3.9.3-5 and Note 4 in Tables 3.4.3-6, 3.9.3-7, and 3.4.3-8 relate to these recommendations.

===Response===
The dy namic loads are combined by the square-root-sum-of the-squares. This is consistent with industry practice and NUREG-0484.
Dynamic events are postulated initiating events or consequential events that contribute to the design basis mechanical load 3 for systems, structures, and components. These events result from starting and stopping of pumps, changes in valse position. pipe breaks, and steam / water interactions. Each fluid system is evaluated against the design basis initiating events to determine if an initiating event is a dynamic event for that particular Guid system, or whether the initiating event causes a dynamic event in that particular Huid system. This evaluation determines whether an initiating or consequential dynamic event can be identified for a particular system for an initiating event. The following screening criteria are used to identify the significant dy namic events:
Pumps:
Pump starts and stops resulting from an initiating esent, or postulated to occur during an initiating event. shall be identified as consequential dy namic events.
Valves:
Fast-opening or fast-closing valves that change position shall be identihed. Fast-opening and fast-closing valves are defined as: remotely-operated salve with a minimum total stroke time of 10 seconds or less; spring-loaded relief s alve; check valve. These events shall be identified as initiating or consequential dynamic events.
Pipe lireaks:
High-energy pipe breaks, for those lines not qualified to Leak-Before-Break criteria, shall be identified. For ASME Class 1, 2, and 3 and seismically analyzed B31.1 piping, these events are postulated as initiating ever.ts only. For non-seismically analyzed B31.1 piping, these events are postulated as initiating or consequential dynamic events.
Stearn/ Water Interactions:
Potential interaction or mixing of cold water and hot steam in a line or component shall be evaluated for its potential to caose a dynamic load. This does not include actuation of pressurizer spray, Dynamic loads that are expected to result from an initiating event are considered in combination with the loads resulting from that event, depending on time phasing of the consequential event and initiating event. Loads resulting from dynamic esents will only be combined with loads resulting from an initiating event if the loads can mechanistically and realistically occur simultaneously.
W Westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION Consequential dynamic loads from an SSE w;ll be combined with SSE depending on the time phasing of the consequential event and the SSE. SSF bads are combined with consequential dynamic loads that can occur as a result of a sinele pipe break in a rsnseismically analy zed piping sy stem.
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SSAR Revision: NONE                                                                                         I i
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210.G3-2 W
Westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION Question 210.64 A
If an clastic-plastic method of analysis will tie used in the design of any safety-related sy stem, component, or support, identity each applicable item and revise either Section 3.9.1 or 3.9.3 of the SSAR to provide information consistent with the guidelines in Section 3. 9. 1.11. 4 of the SRP.

===Response===
for systems w here service lesel D limits are specified for safety-related piping and supports, the method of anal) sis used to calculate the stresses and deformations shall conform to the methods outlined in ASN1E Section ill, Subsection NF and Appendix F. He inelastic analysis criteria included in Appendix F will be used as an alternative to the procedures of NB-3652 of the ASN1E Section ill code when the inelastic analysis option will provide a more cost effective design. No particular system, component, or support is presently identified for this ty pe of design evaluation.
When an clastic analysis is performed for SSE and an inelastic anal) sis is performed for pipe break loadings, the following conditions will be satisfied:
: a. The stress-strain relationship for the actual ty pe of material undergoing plastic deformation will be used in the analysis.
: b. The ultimate strength value at service temperature is not important for this evaluation, because only small strains ( < < ultimate strain) will be permitted.
: c. The analy tical procedures used in the inelastic analy sis will be those associated with the Westinghouse proprietary code, WECAN.
: d. The applicability and validity of the WECAN code for use in inebstic analysis will be provided to meet the requirements of 10 CFR Part 50, Appendix B and GDC 1.
: e. The appropriate interaction of clastic and inelastic components will be demonstrated to provide assurance that system displacements and support deformations do not violate assumptions on which the system analysis is based. This should not be a problem because only piping will be allowed to go plastic, and therefore, the sy stem displacements are the analytically correct displacements.
: f. The support load results from an inelastic analysis for Design Basis Pipe Break (DHPH) will be combined with SSE results using Table 3.9-8.
: g. The piping stress results from an inelastic analysis will be combined as stated in Tables 3.9-6 and 3.9-7 for the elastic piping elements. For the plastic piping elements, the element strain associated with the SSE condition will be added absolutely with the strain associated with the DBPB condition. The strains will be limited to 1%
for strains aseraged through the thickness; 2% for strains at the surface, due to an equivalent linear distribution of strain through the thickness; and 5% for local strains at any point.
When an inelastic analysis is performed for SSE and pipe break loadings the method in Table 3.9-6. note 19 is used.
See response to RAI 210.68 for revisions to Tables 3.9 9 and 3.410. See response to RAI 210.74 for resisions to Tables 3.4-6. 3.9-7, and 3.9-8.
210.64-1 W-Westinghouse

4 .
NRC REQUEST FOR ADDITIONAL INFORMATION                     l
    , . - - - . -                                                                                                            I A
S5AR Revision:
3 A 3.1.5       ASNIE Classes 1. 2, and 3 piping The loads for ASN1E Code Classes 1. 2. and 3 piping are listed in Tables 3.9-3 and 3.9-4. Tables 3.9-6 and 3.9-7 lists the loading combinations. Tsbles 3.9-9, 3.9-10, and 3.9-11 presents the stress limits.
Piping s) stems are designed and an.Jyzed for Ixvels A, B. and C service conditions, and corresponding service level requirements to the rules of the ASME Code, Section III. The anal) sis or test methods and associated stress or load allowable limits that are used in evaluation of 12sel D service conditions are those that are defined in App (ndix F of the ASME Code, Section III.
Subsection 3.7.3. summarizes seismic analysis methods and criteria. Subsection 3.6.2 summarizes pipe break analysis methods.
The supports are represented by stiffness matrices in the system model for the dynamic analysis. Alternate methods for support stiffnesses representation is provided in Subsection 3.9.3.4. Shock suppressors that resist rapid motions are also included in the analysis. The p2 solution for the seismic disturbance uses the response spectra method. This method uses the lumped mass technique, linear elastic properties, and the principle of modal superposition.
The total response obtained from the seismic analysis consists of two parts: the inertia response of the piping system and the response from differential anchor motions. (See Subsection 3.7.3). The stresses resulting from the anchor motions are considered to be secondary and are evaluated to the limits in Table 3.9-11.
The mathematical models used in the seismic analyses of the Class I, 2. and 3 piping systems lines are also used for pipe rupture ef fect analysis. To obtain the dynamic solution for auxiliary lines with active valves, the time-history deflections from the analysis of the reactor coolant loop are applied at nozzle connections. For other lines that must maintain structural integrity or that have no active valves, the motion of the reactor coolant loop is applied statically.
When an elastic analysis is performed for SSE and an inelastic analysis is performed for pipe break loadings, the following conditions are satisfied:
: a. The stress-strain relationship for the actual type of material undergoing plastic deformation is used in the analysis,
: b. The ultimate strength value at service temperature is not important for this evaluation, because only small strains (< < ultimar.e strain) will be permitted. The ultimate strength value is obtained from the ASME III Code, Appendit I..
: c. The analytical procedures used in the inelastic analysis are those associated with the WECAN computer code.
: d. The applicability and validity of the WECAN code for use in inelastic analysis meet the requirements of 10 CFR Part 50, Appendix B and GDC 1.
: e. The appropriate interaction of elastic and inelastic components will be demonstrated to provide assurance that system displacements and support deformations do not violate assumptions on which the system analysis is based. This is accomplished by allowing only the piping to go plastic and maintaining pipe supports as clastic elements that meet the amits of ASME Section III, Subsection NF and Appendix F. Therefore, the system displacements are the analytically correct displacements.
: f. The support load results for DBPB are combined with SSE results using Table 3.9-8. for either elastic or inelastic system analysis.
210.64-2 W-W85tlngh00SO

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NRC REQUEST FOR ADDITIONAL INFORMATION                                                                               l y
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: g. The piping stress results from an inelastic analysis are combined as stated in Tables 3,9-6 and 3.9-7 for the elastic piping eternents. For the plastic piping elements, the element strain associated with the SSE condition are added absolutely with the strain associated with the Dl3PB condition. The strains are limited to 1% for strains averaged through the thickness; 2% for strains at the surface, due to an equivalent linear distribution of strain through the thickness; and 5% for local strains at any point.
When an inelastic analysis is performed for SSE and pipe break loadings the method in Table 3.9-6, note 19 is used.
A thermal transient heat transfer analysis is performed for each different piping component on the Class I branch lines larger than l-inch nominal diameter. The following discussion on the esaluation of cyclic fatigue is not applicable to Class 2 and 3 pipe.
The inel A and H sersice condition and test condition transients identified in Subsection 3.9.1.1 are included in the fatigue es aluation. For each thermal transient, two load-sets are defined representing the maximum and mini-mum stress states for that transient.
The primary-plus-secondary and peak stress intensity ranges, fatigue reduction factors, and cumulative usage factors are calculated for the possible load-set combinations. It is conservatively assumed that the transients can occur in any sequence, thus resulting in the most conservative and restrictive combinations of transients.
The combination of load-sets yielding the highest alternating stress intensity range is determmed, and the incremental usage factor is calculated. Likewise, the next most severe combination is then determined, and the incremental usage factor is calculated. This procedure is repeated until the combinations having an allowable cycle of less than 1011 are formed. The total cumulatise usage factor at a point is the summation of the incremental usage factors.
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i 210.64-3    l Waus Westinghouse                                                                                                  ,
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NRC REQUEST FOR ADDITIONAL INFORMATION Question 210.68 Section 3.9 3.4 of the SSAR does not appear to specifically address allowable stress criteria for actise component supports, w here active is as defined in Section 3.9.3.2.2. The staff's position is that the stresses and associated deformations in such supports should be low enough to allow operability of the supported component, in Appendix I A of the SSAR and Revision I to WCAP-13054 (under exceptions to Section 3.9.3 of the SRP),
exceptions are taken to Position C8 in RG 1.124. and Paragraph B.5 in RG 1.30 which are the bas, s for the staff's position on this issue. The exceptions in Appendix 1 A state that ASME Level C and D Service Limits are acceptable. however, when they are used, any significant deformation that might occur will be considered in the evaluation of equipment operability. Revise Section 3.9.3.4 to reference this exception and provide a more detailed discussion on how this significant deformation will be evaluated for the AP600 to meet the guidelines in Section 3 . 9. 3 .11. 3 of the SRP. Appropriate revisions should also be made to Tables 3.9-9 and 3.9-10, and Appendix ! A of the SSAR, and to the exception to Section 3.9.2.ll.3.a of the SRP in WCAP-13054.

===Response===
In order to assure operability for active equipment, including valves, ASME limits for Service Level C loadings will be met for the supports.There will not be significant support deflections for these stress levels since only the outer fibers are permitted to yield.
For pipe supports on lines with active valves the support in the sicinity of the vab es will meet either of the following to assure that support deflections are not significant:
a) ASME limits for Senice level C loadings b) Inelastically calculated support displacement will not exceed 1/8 inches Stress criteria for piping is shown in Tables 3.9-6 and 3.9-7(see response to question 210.79)
WCAP 13054 will be revised to delete exception to SRP 3.9 3.II.3.a.
SSAR Revisions:
3.9.3.4 Component and Piping Supports The supparts for ASM'l Code Section 111. Class 1. 2. and 3 components including pipe supports satisfy the requirements of th- ASME Code. Section 111. Subsection NF. The boundary between the supports and the building structure is based on the rules found in Subsection NF. Tables 3.9 3 and 3.9-4 present the loading conditions.
Table 3.4-8 summarizes the load combinations. The stress limits are presented in Table 3.9-9 and 3.9-10 for the sarious service levels.
The criteria of Appendix F of the ASME Code Section 111 is used for the evaluation of Level D senice condi-tions. When supports for components not built to ASME Code. Section 111 criteria are evaluated for the effect of l
1 W Westinghouse 1
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NRC REQUEST FOR ADDITIONAL INFORMATION A
Incl D service conditions, the allow able stress levels are based on tests or accepted industry standards comparable to those in Appendix F of ASME Code, Section 111.
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In order to provide for operability of active equipment, including valves, ASME limits for Service level C loadings will be met for the supports of these items. When elastic system analysis is used for pipe supports on lines with active valves the supports in the vicinity of the valves will meet either of the following to provide that deflections do not have a significant effect on the elastic piping system analysis, a) ASME limits for Service level C loadings b) Inelastically calculated support displacement will not exceed 1/8 inches The vicinity is defined as supports that are within a distance from the valve of one-half the standard deadweight span in ASME 111, NF.
If an elastic system analysis is not able to satisfy (a) or (b) above to ensure operability, then an inelastic system analysis will be perfonned which accounts for the inelastic behavior of the piping and supports. This analysis is performed in such a manner that valves and equipment, which are required for operability, remain clastic and satisfies other operability requirements.
Dynamic loads for components loaded in the elastic range are calculated using dynamic load factors, time-history analysis, or any other method that accounts for elastic behavior of the component. A component is assumed to be in the elastic range if yielding across a section does not occur. Local yielding due to stress concentration is assumed not to affect the validity of the assumptions of elastic behavior. The stress allowahles of Appendix F for elastically analyzed components are used for Code components.                                                                ;
Dy namic loads for component supports loaded in the inelastic range are calculated using dynamic load factors.
time history analysis, or other methods that account for the inelastic behavior. The analysis requirements of paragraph F-1322 of Appendix F of the ASME Code. Section ill are satisfied.
The stiffness of the pipe support miscellaneous steel is controlled by one of the following methods so that component noule loads are not adversely affected by support deformation:
Pipe support miscellaneous steel deflections are limited for dynamic loading to 1/8 inch in each restrained direction. These deflections are defined with respect to the structure to which the miscellaneous steel is attached.
These deflection limits, provide adequate stiffness for seismic analysis and are small enough so that nonle loads are not affected by pipe support deformatior.. In this case, the pipe support and miscellaneous steel are represented by a generic stiffness value in the piping system analysis. Rigid stiffness salues are used for fabricated supports.         t and vendor stiffness values are used for standard supports such as snubbers, rigid gapped supports, and energy-              i absorbing supports. The mass of the pipe support miscellaneous steel is evaluated as a self weight excitation loading        j on the steel and the structures supporting the steel.                                                                        ji l
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210.68-2 W
Westinghouse            l l

Table 3.9 9                                                                        2 m
i                                          Stress Criteria for ASME Code Section ill o
m m
g                         Class 1 Components (a) and Supports and Class CS Core Supports                                                      o C
                                       !a                                                                                                                                             m 5                                                                                                                                              N Design' Service                                                                                                                      y I;evel      Vessels' Tanks                                 Core                                                                 m Pumps              Piping             Supports      Valves. Disk &   Components                              g Seats       Supports (C)(d)                       o Design and ser-    A S AI E   Code. M414'rC-ede          A S NI E   Code. A S NI E    Code. A S NI E    Code,                      d vice level A       Section Ill N B-   Fe.' en !ll- N 4 Section ill NG-      Section ill N B-    Section til Sub-                      @
3221,3222          A52, 'M2      See    3221.3222.3231,  3520, 3525         section NI: NF-                        >
Table 3.9-6          3232                                122!. 3:22 N F-                        h 4141+a&NM240 O
(e)                                   m Sen ice level B    A S hl E   Code. M4I F C :de          A S NI E   Code, A S hl E     Code. A S NI E    Code.                     3 (Upset)            Section ill N H-   Sect en lil-N4 Section 111 NG-        Section ill N B-   Section 111 Sub-                      $
3223                As4 See Table        3223, 3233       3525               section NF (e)                        O 3.9-6                                                                                          2 NM223 323!.
4M-NM240 Senice level C     A SSI E    Code. A#M F       Cede. A S hl E Code.      A S NI E     Code. A S hl E    Code,
( Emergency       Section til N B-    Sc.
* en !Il-N4 Section Ill NG-       Section III N B-   Section Ill Sub-3224                3'55 See Table 3224,3324              3526               section NF (e) 3.9-6                                                    NIL 322! 3231-4 (4 % M 240 Sen ice les el D A S hl E Code, A&M F Code A S hl E Code.                   (b)                A S NI E    Code.
(Faulted)         Section 111 (see } S c' en !Il- See Section ill (see i                       Section 111 Sub-3.9.1.4) NB-3225 Table 3.9-6 (see i     3.9.1) NG-3225                       section NF (e).
(No active Class 1  3.9.1.4HNH 36 55     3335                                (see p 3.9.1) NIL pumps used)                                                                  3225      3211-4M NIL 1240 (f)
N a                                                                                                                          ---
P.
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9 to                                                                                                                        m

NRC REQUEST FOR ADDITIONAL INFORMATION A                Notes to Table 3.9-9:
: a. A test of the components may be performed in lieu of analysis,
: b. Class 1 valve Service Level D criteria for active valves and inactive valves is based on the criteria in ASN1E 111 Appendix F, F-1420 for veri 6 cation of pressure boundary integrity. Valve operability is demonstrated by testing.
: c. Including pipe supports.
: d. In instances where the determination of allowable stress values utilizes Su (ultimate tensile stress) at temperatures not included in ASME Code Section ill, S ushall be calculated using one of the methods provided in Regulatory Guide 1.124. Revision 1.
: e. ASME Table 3131(a)-l.
: f. See subsection 3.9.3.4 for supports for active equipment, valves, and piping with active valves.
210.68-4 W
Westinghouse

Table 3.9-10                                                       2 x
I                                                                                                                         o Stress Criteria for ASME Code Section 111                                             m m
g                                  Class 2 and 3 Components and Supports                                                 o C
C g      Design /Sen ice   Vessels / Tanks        Piping              Pumps      Valves. Disks,  Component g          inel                                                                   Seats       Supportst aHb)             g E
                                                                                                                          'D Design and      AShlE Code A&M F Cnde                   A S.\t E Code  ash 1 E Code     ASNIE Code                 g senice level A   Section 11I        F rc ti :n    !I4 Section ill       Section      111 Section Ill(c)             g NC-32I7 NC4W 2 ^ i2 -                   N C/N D-3*X)   NC/N D-3510      N M 124                    -!
NC/N D-33 to,      -4 M See Table                                       N M 244                    @
3320               3.9-7                                                NMN)                       >
r-Cah AShlE Code E
Service lesel B AShlE Code AKMr                                         AShl E Code      AShlE Code                  n (Upset)          Section ill S+++4**--14 4 Section 111                  Section     lli  Section Ill(c)             O N C / N D-3310,    NC! N        D--     NC/ N D-3*K)   NC/N D-3520      NM124                      E 3320               M ISee Table                                         NM244                      M 3.9-7                                                NMN)                       6 2
Service level C  AShlE Code         A&MF Cnde           ASNIE Code AShlE Code            AShlE Code
( Emergency)     Section ill See4 W il-                 Section     ill Section     ill  Section Ill(c)
N C/ N D-3310      NC M5 ?See          NC/N D-34N)     NC/N D-3520      NW1 3320               Table 3.9-7                                          NW NW1 Service level D  ASNIE Code A&M F Cnd                   AShlE Code AShl E Code           ASNIE Code (Faulted)        Section III        Sce' en       !!1   Section ill     Section 111      Section lil(c)(d)
N C/ D-3 310,      NC 3 ^ < 5 S e e    NC/N D-3400     NC/N D-3520      NR4424 3320               Table 3.9-7                                          NF-1114 NM244)
N o'                                                                                                            ,
o 03                                                                                                                 :::
: 6.                                                                                                             - all

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NRC REQUEST FOR ADDITIONAL INFORMATION l
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       'A l                                  Notes to Table 3.9-10:
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: a. Including pipe supports,
: b. In instances where the determination of allowable stress values utilizes Su (ultimate tensile stress) at I          temperatures not included in ASME Code Section III, Sushall be calculated using one of the methods provided in Regulatory Guide 1.124, Revision 1.
: c. ASME Table 313Ha)-l.
l      d. See subsection 3.9.3.4 for supports for active equipment, valves, and piping with active valves.                  i l
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210.G8-6 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 210.69 l
Section 3.9.3.4.3 of the SSAR does not provide suf ficient information for the staff to conclude that snubber            j operability will be assured. Revise this section to provide a more detailed discussion which incorporates the            I guidelines in Section 3.9.3.11.3 of the SRP. In addition, if applicable, provide a commitment to dynamically qualify     l all large bore bydraulic snubbers.                                                                                       l l
l The discussion of Generic Safety Issue A-13,
* Snubber Operability Assurance " in Section 1.9.4.2 of the SSAR should also be revised to reference the revised Section 3.9.3.4.3.                                                       l l

===Response===
The SSAR Subsection 3.9.3.4.3 will be revised as shown below to address the SRP guidelines and provide a more detailed discussion of the production and qualification tests. The discussion of Generic Safety issue A-13. Snubber Operability Assurance, in Subsection 1.9.4.2 of the SSAR already references Subsection 3.9.3.4.3.
SSAR Revision:
3.9.3.4.3       Snubbers Used as Component and Piping Supports The location and size of the snubbers are determined by stress analysis. Access for the testing, inspection, and maintenance of snubbers is considered in the AP600 layout. The location and line of action of a snubber are selected based on the necessity of limiting seismic stresses in the piping and nozzle loads on equipment. Snubbers are chosen in lieu of rigid supports where restricting thermal growth would induce excessise thermal stresses in the piping or nozzle loads or equipment. Snubbers that are designed to lock up at a given velocity are specified with lock-up velocities sufficiently large to envelope the highest thermal growth rates of the pipe or equipment for design thermal transients. The snubbers are constructed to ASME Code, Section III, Subsection NF standards.
In the piping sy stem seismic stress analysis, the snubbers are modeled as stiffness elements. The stiffness value is based on vendor stiffness data for the smtbber, snubber extension, and pipe ehmp assembly. Supports for actise valves are included in the overall design and qualification of the valve.
j The elimination of the analysis of dynamic effects of pipe breaks due to leak-before-break considerations, as outlined in Subsection 3.6.3, permits the use of few er snubbers than in plants that were designed without considering leak before break. Also, the AP600 design makes use of gapped support devices to minimize the use of snubbers.
The evaluation of those snubbers used as supports is outlined below.
i                              Design specifications for snubbers include:
* Seismic requirements
* Normal environmental parameters
* Aceident' post-accident environmental parameters
* Full-scale perfonnance test to measure pertinent perfonnance requirements
* Instructions for periodic maintenance (in technical manuals)
Two types of tests are performed on the snubber to verify proper operation.
W85tlngt100S8

NRC REQUEST FOR ADDITIONAL INFORMATION u    :;pg I
* Production tests on every unit to verify proper operability i
Qualification tests on randcmly selected production models to demonstrate the required load performance (load                      j rating)                                                                                                                            i The penduction operability tests for large hydraulic snubbers (i.e., those with capacities of 1000 kips or greater)                    l include a) a full Level D load test to verify sufficient load capacity, b) testing at ftdl load to verify proper bleed with the control valve closed, c) testing to verify the control valve closes within the specified velocity range, and d)                    )
testing to demonstrate that breakaway and drag loads are within the design limits.
* The operability of essential snubbers is verified by *.he COL holder by verifying the proper installation of tb l
,     snubbers, and performing visual inspections and measurements of the cold and hot positions of the snubbers as                          -
required during plant heatup to verify the snubbers are performing as intendext.                                                       i b
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NRC REQUEST FOR ADDITIONAL INFORMATION 1
A Question 210.88 Resision I to WCAP-13054 lists an exception to Section Se of Section 3.10 of the SRP. that states that Westinghouse does not prepare a Seismic Qualification Report (SQR), and that, in lieu of such a report seismic qualification of equipment is documented in test reports, analysis reports. calculation notes, etc. contained in Westinghouse files. The staff's position is that an SQR should be prepared, and included in the documentation prosided by the COL (see Q210.86). Resise this exception to state that the SQR should be submitted by the COL applicant.
In addition verify the existence of design and analysis documentations of reactor internals, and provide a summary of the analysis results in conjunction with design limits.

===Response===
The information recommended to be included in a Seismic Qualification Report is included as part of an equipment qualification data package for each piece of equipment considered. Details of the documentation process are described in AP600 SSAR Section 3.10 and Section 3D.7 (Appendix 3D). The Combined License applicant will maintain this information as the equipment is selected and procured. The data packages will be as ailable for review and audit by the NRC.
The design and analpis documentation for the reactor internals will be available as part of the ASME Code Design Report w hen construction of the reactor internals is complete.
The position for Section $c of Section 3.10 of the Standard Review Plan will be revised in the next revision of WC A p- 13054.
The serification of the existence of design and analysis documentation of reactor internals and a summary of the analysis results in conjunction with design limits is provided as an attachment to this sesponse for information. The summary includes the components of reactor internals which are not similar or can not he enveloped by previously designed Westinghouse plants. See attachment to this RAI for the summary of stresses.
SSAR Revision:
See the respon:,e to RAI 210.86 for the SSAR Revision.
0.884 W-Westingtiouse

l NRC REQUEST FOR ADDITIONAL INFORMATION ATTACHN1ENT TO RESPONSE TO RAI 210.88 SUN 1N1ARY OF REACTOR INTERNALS ANALYSIS RESULTS R ADI AL KEY AND CLEVIS INSERT N1 ARGINS OF SAFETY (I) TAllLE Section                         Load Category         Stress Type        Stargin of Safety    Fatigue Usage Key liase                       Level A + Il          Pm                0.18 Pm + Pb           0.195 Pm + Pb + Q       Large
* EU                     Fatigue          ---
0.026 < l.0 Allow able level D               Pm                0.265 Key Itase 45                    level A+ B            Pm                0.013 Key Bearing                     Level A+ B            Pm                0.227 Weld                            level A + 11          Pm                0.77 Pm + Pb + Q       0.203 EU                    Fatigue           ----
0.449 < l.0 Allowable level D               Pm                 1.25 Clesis insert Flange            level A + 11          Pm                0.074 Pm + Pb            >0.074 EU                    Fatigue           ----
Approx.0.00 <
l.0 Allowable Clevis insert                   les el A + 11         Pm                0.29 Pm + Pb           Large*
EU                    Fatigue           ----
0.80 < l .0 Allowable lesel D               Pm                0.326 Note : (1) .\largin of Safety = Allowable Stress Intensity / Calculated Stress Intensity - 1
        = Greater than 10 210.88-2 3 Westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION r..   ...
Section              Load Category  Stress Type  Margin of Safety Fatigue Usage Clesis Fasteners     level A+ B     Pm           1.52 (Bearing) 4.52 (Shear)
Pm + Pb     Large*
Pm + Pb + Q 0.816 EU             Fatigue     - - - -
Approx. 0.00 <
l.0 Allowable Dow el Pin            level A + B   Pm          Large*
Pm + Pb     Large*
Pm + Ph + Q  l.28 EU              Fatigue     ----
Approx. 0.00 <
l.00
     * = Greater than 10 W Westinghouse                                                            210.88-3

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NRC REQUEST FOR ADDITIONAL INFORMATION I
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UPPLR SUPPORT ASSEMBLY MARGIN OF SAFETY TABLE l
l      Section                       Load Category                Stress Type                            Stargin of Safety Fatigue Usage     !
Perforated Region             Ie.cl A + fl                 Pm                                     l arge*
Pm + Pb                                1.746                                l Pm + Pb + Q                            l.55 l
EU                           Fatigue                               ---
0.0013 < l.00 Allowable lesel D                      Pm                                    2.12 Pm+ Pb                                3.67 Skirt / Flange                lesel A + B                  Pm                                     12rge*                               l l                                                                 Pm + Pb                               0.81                                  l l
l                                    Level D                      Pm                                     12.59                                l Pm + Pb                                12.67 l      Skirt / Plate                 Level A + B                  Pm                                    8.0                                   ,
1 level A + B Upper Support Flange                                       Pm                                    4.77 l                                                                 Pm + Pb                               3.42 Pm + Pb + Q-                          3.52 EU                            Fatigue                               ----
Approx. 0.00 <
l.00 Allowable Level D                       Pm                                    2.22 Pm + Pb                               3.12
       * = Greater than 10 210.88-4 W85tlngh00S8

NRC REQUEST FOR ADDITIONAL INFORMATION                                                                             j l
l Section              Load Category  Stress Type Margin of Safety         Fatigue Usage Skirt to USP Weld    Level A +13    Pm          7.24 Pm + Pb     4.62 Pm + Pb + Q 0.22 EU             Fatigue     ----
                                                                              .0143 < l .00 Allowable level D        Pm          4.14 Pm + Pb     2.89 Outer G/T location   Steady State   Rotation,   0.00131 <
0, rad      0.0016 i
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210.88.s                         I W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION LOWER CORE SUPPORT PLATE ASSEN111LY SIARGIN OF SAFETY TABLE Section                  Load Category                        Stress Ty pe                                                        Stargin of Safety                                    Fatigue Usage Inner Perforated Region  level A + B                          Pm                                                                  large
* Pm + Pb                                                            0.25 EU                                   Fatigue                                                            ----
0.255 < l.00 Allow able level D                              Pm                                                                 Large*
Pm + Pb                                                            0.41 Transition Rim Region    InelA+B                                 'm                                                               10.7 l                                                                    Pm+Pb                                                              2.9 EU                                   Fatigue                                                            - - - -
0.299 < l.00 Allowable 12 vel D                             Pm                                                                  16.0 l                                                                    Pm+Ph                                                               12.0 Level D                              Pm                                                                  i1.37 Pm + Pb                                                             17.56 Center 1.CP              Steady state                         De fl. , 5. inch                                                  0.032 < 0.060
       * = Greater than 10                                                                                                                                                                                                      l I
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__                                                         _ . _ _ _ _ _ _ _ _ _ _ . _ _ . . . _ _                                                                  . _ _ _ _ _      ________________________.__m

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NRC REQUEST FOR ADDITIONAL INFORMATION A
VORTEX RING ASSEMllLY M ARGIN OF SAFETY TAllL.E Section                                              Load Category                                                                          Stress T.5 pe                            Margin of Safety Fatigue Usage Vortex Ring                                           12sel A + B                                                                           Pm                                       large*
Pm + Pb                                  21.6 Pm + Pb + Q                              0.4457 EU                                                                                      Fatigue                                 -----
0.0017 < l.00 Allowable level D                                                                                Pm                                       large*
1 Pm + Pb                                  5.9                                         l l
Sec. Core Supt Col.                                  12 vel A + B                                                                           Pm                                       81.7 Pm + Pb                                  4.23 EU                                                                                      Fatigue                                 -----
0.018 < l .00 Allowable lesel D                                                                                Pm                                       41.8 1
Pm + Pb                                   1.11                                       l Core Drop                                            Level D                                                                                Buckling                                 0.206 Inner Column Flange                                  level A + B                                                                            Pm                                       0.084 I                                                                                                                                                  Pm + Pb                                  0.352 l
Pm + Ph + Q                              3.15 EU                                                                                      Fatigue                                 ----
Approx.0.00 <
l.00 Allowable level D                                                                               Pm                                       4.8                                         l Pm + Pb                                  6.29
       * = Greater than 10 1
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NRC REQUEST FOR ADDITIONAL INFORMATION CORE B ARREL (UPPER AND LOWER) M ARGIN OF SAFETY TABLE Section                           Load Category   Stress Ty pe                              Margin of Safety                   Fatigue Usage l
lower Core Barrel                 level A + B     Pm                                         1.03 Pm + Pb                                  2.04 l                                        EU               Fatigue                                 ----
0.03 < l .00 Allow able Upper Core Barrel                 level A+ B      Pm                                       0.67 Pm + Pb                                  0.92 EU               Fatigue                                 - - - - -
0.095 < l .00  ;
Allowable     j lesel D         Pm                                         1.51 i                                                        Pm + Pb                                  2.44 1
Core Barrel Flange                lesel A + B     Pm                                       0.87 Pm + Pb                                  0.04                                              ,
i EU               Fatigue                                                                    0.029 < l.00 Allowable      j i
Lesel D         Pm                                         1.26 Pm + Pb                                  0.53 l
!      Core Barrel Noi.zle               level A + B     Pm                                       35 l
l                                        EU               Fatigue                                 ----
0.894 < l .00 Allowable Level D         Pm                                       41 210.88-8 W
                                                                                                                                    - Westin ause

NRC REQUEST FOR ADDITIONAL INFORMATION REFLECTOR      ASSEMilLY Section               Load Category   Stress Ty pe     Margin of Safety  Fatigue Usage Rottom Block          le sel A + B    Pm               > 1.18 Pm + Pb          > 1.18 Usage Factor    Fatigue         - - - - -
0.044 < l .00 Allow able level D         Pm               > 1.62 Pm + Pb          > l.62 Center 14 >ck        level A + B      Pm               > l.18 Pm + Pb          > l.18 Usage I actor    i atigue       -----
0.014 < l.00 Allowable level D          Pm               > 1.62 Pm + Pb          > l.62 Top tilock           Level A + B      Pm              > 1.18 Pm + Pb         > 1.18 Usage Factor     Fatigue        - - - -
0.156 < l.00 Allowable      l Top Block            level D          Pm              > l.62 Pm + Pb         > 1.62                           )
Top Flange            level A + B      Pm              > 1.18 Pm + Pb         > 1.18 Usage Factor     Fatigue        -----
0.156 < l.00 Allowable level D          Pm              > 1.62 Pm + Pb         > l.62
[ W85tingh0USB                                                                       '

NRC REQUEST FOR ADDITIONAL INFORMATION A
Tie Rod       level A + B  Pm           4.9 Usage Factor Fatigue      -----
0.0016 < l.00 Allow able le vel D     Pm           0.45 Alignment Bar level A + B  Pm           6.39 Pm + Pb      2.71 Usage Factor Fatigue      ----
0.013 < l.00 Allowable level D      Pm           0.32 Pm + Pb      7.86 210.88 10 W Westinghause

l NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 l
l     Quesuco 440.1 l
l     a. Provide updated or revised topical reports on the AIW)0 test progr:un, including WCAP-13277, " Scaling, Design, and Verification of the SPES-2, the Italian Experimental Facility Simulator of the AP600 Plant,"
and WCAP-13234. "AP600 Long Tenn Cooling Test Specification." An accurate representation of the facility design, a scaling analysis reflecting that design, a detailed test matrix, and an analysis plan should be provided in the reports.
: b. Provide topical reports detailing planned testing for the long-term check valve testing program and the departure from nucleate boiling testing program. Indicate whether the " biased-open" check valves will be tested and,if so, a topical report detailing the test specification should be provided. If the " biased-open" check valves are not to be tested, provide a detailed explanation why such testing is not required.
: c. Provide WCAP-12980, "AIW)O Passive Residual Heat Exchanger Test Final Report," that is referenced in the SSAR.
1 Response (Revision 1):
: a. WCAP 13277, " Scaling, Design, and Verification of SPES-2, the Italian Experimental Facility Simulator          ,
of the AP600 Plant " Revision 1, was pruvided to the NRC via Westinghouse letter ET-NRC-93-3883, dated May 11,1993.
l l           WCAP 13234, "AP600 Long Term Cooling Test Specification". Revision 1, was provided to the NRC via Westinghouse lette ET-NRC-93-3883, dated May 11,1993.
7 WCAF4M&and-l W4-will4w-updated-anJ46wwarded4Wanuaryr49%
1 Check Valve Test Documentation b.
l A4opicalaepwkletailing4he-planned 4esting4or4he " keg 4crm"-ched ' ""e 'estinfHweg=' '= -~ : et been-petwedr-We.tingho+se4w-initiated : s :e :4-enisting-+aility-inf.e= m * :e.= diedr. valve opening-performanee-after4eing-elosc4aM+iglaleha-P4or-a4ong4imeri+.-eendith*w+imilaF ! thme Sh!ch would4%ngerie#weJ4>y-4Wravity-dr ' "" "%es in = der te :ermw4 hat 4he L : ;trogranWit,aekirew releVte64-fMt4% Westinghouse is assessing check valve opening perfonnance for valves that have been closed at high delta P for a long tirne, These tests are being performed at existing PWR sites during refueling outages. The report on the first of these in-situ tests is documented in WCAP-14045, "In Situ Check Valve Test Report (Fall Outage '93)," which was transmitted to the NRC via Westinghouse letter NTD-NRC-94-4120, dated May 4,1994.
             " Biased-open" check valves will not be tested in this program. These valves are open and are not expased to high differential pressure.
W WestinEhouse

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l NRC REQUEST FOR ADDITIONAL INFORMATION l
i!!" "lii Response Revision 1 Additional infonnation on check valve testing is provided in WCAP-13560. " Advanced Plant Check Valve Study" Rev. O. which was provided to the NRC in letter ET-NRC-93-3801, "AP600 Testing 1(eports (WCAP- 13567, WCAP-13566 and WCAP-13560), from N. J. Liparulo to Dr. Thomas Murley, January 25, 1993.
Infornationm4)ND -Testing-wiWprovide44nJanuary-144 DNB Test Docurnentation DNB testing was performed in acconiance with WCAP 12488, " Westinghouse Fuel Criteria Evaluation Process" which was provided to the NRC as a topical report for review under letter NS-NRC-90-3482, to V. H. Wilson from W. J. Johnson, April 2,1990. Further infonnation has been provided in the following submittals:
ET.NRC-92-3702, " Responses to Request for Additional Infonnation on WCAP-12488,
                      " Westinghouse Fuel Criteria Evaluation Process", from N. J. Liparulo to R. C. Jones, June 8,1992.
ET-NRC-92-3723 " Supplement to AdditionalInfonnation on WCAP-12488," Westinghouse Fuel Criteria Evaluation Process" frorn N. J. Lipamlo to R. C, Jones, July 17,1992.
ET-NRC-93-3819, " Final Responses to Additional Information on WCAP-12488, " Westinghouse Fuel Criteria Evaluation Process", from N. J. Liparulo to R. C. Jones, February 8,1993.
ET-NRC-93-3984, " Test Matrix for AP600 Depmture from Nucleate Boiling Tests," from N. J.
Liparuto to R. W. Borchanit. October 8,1993.
The final test report on the AP600 DNBR tests will be transmitted to the NRC in August 1994.
: c.       PRHR Test Documentation WCAP-12980, AP600 PRHR HX Test Final Report", was provided to the NRC in letter ET NRC-92-3779,
             " Submittal of Al%00 Design Certification Material required by 10 CFR 52.47", from N. J. Liparulo to Dr. Thomas Murley. December 15.1992. "AP4KAPe.iwRewlaaLHe+4 Fashanger Test FiraLRm*1h"-wiH 4+4orwarded4w4mmayAv%
SSAR Revision: NONE l
440.1(RI)-2 W*
Westinghouse        ,
1

NRC REQUEST FOR ADDITIONAL INFORMATION
                                                                                                              .hb   ET$hh Response Revision 1
* 5 Question 440.11 Revision O of WCAP-13342, "AP600 Automatic Depressurization System Test," is dated January 1991. An updated version of this test specification should be provided, incorporating any changes in the design or test plans for the test anities in Phase A and Phase B, particularly as a result of changes in the AP600 plant design.
Response (Revision 1):
WCA P-l M42rA FVMAutomati+4Mres .urirati++Svende44*+urrently4+ing*f wlatedr+j wikallvanimqwaie a(kli t ional w4ev i+ Linformat i+ +4or4 he4%+4W41mpropamr-Thi+revb ion-wilfalw4sw4uduhange*4o4 tw Phaw-A4e4-propam 4,4evi+1-WCAP+ ilk %vovided 4n4anuarw-44%
WCAP-14112, Revision 0, "AP600 Automade Depressurizadon System - Test Specificadon (Phase B1)" was provided to the NRC via Westinghouse letter NTD-NRC-94-4178.
WCAP-13891, Revision O. "AP600 Automatic Depressurizadon System Phase A Test Data Report," was provided to the NRC via Westinghouse letter NTD-NRC-94-4176, dated June 20,1994.
SSAR Revision: NONE W WestinEhouse

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NRC REQUEST FOR ADDITIONAL INFORMATION Question 440.52 The staf f has determined that the scaling report on the core makeu tank (CMT) does not provide sufficient information to demonstrate that the CMT separate-effects test will represent the processes occurring in the AP600 plant during operation of the CMTs. In addition, there appear to be inconsistencies in the report. Address the follow ing concerns:
: a. The description of the operation of the CMT described in Section 1-2 of the report appears to be inconsistent with Figure 1-1. On page 4, it states that "the discharge line isolation valves are normally open." This is inconsistent with the drawing referenced in the discussion (Fig.1-1), which shows the s alses closed. In addition, this description appears to be incorrect, since wi h the CMT at reactor coolant system (RCS) pressure, leaving the discharge valves open would establish a circulation path from the reactor vessel through the pressurizer to the CMTs, and the tanks would drain slowly. From the AP600 SSAR, it has been understood that all CMT isolation valses are normally closed (but fail open), and open on the various CMT actuation signals. Clarify this inconsistency.
: b. Clarify the nomenclature in Chapter 2 of the report (from page 141) used for the conservation equations for the CMT. Specifically, in the momentum equation, as show n in Equations 2.2 and 2.7, and in the dimensionless parameters derived therefrom, the terms ac and I appear, which are defined as core area and subchannel length, respectively. While the equation per w is appropriate, insofar as consistency with the stated assumptions (1-D steady-state momentum) is concerned, the reference to core parameters in the CMT equation does not appear appropriate. The symbology or the nomenclature should be corrected.
: c. In Section 2-2 of the report, in the discussion of convective heat transfer during the recirculation mode of CMT operation, it is stated (on page 30) that the Rey nolds numbers for both the plant's CMT and the test article are about 6000-7000, and that this is a " turbulent flow regime such that a turbulent heat transfer consectise correlation such as Ditrus Boelter...isapplicable." The staff concludes that a Re value of 6000-7000 places the system in the laminar-to-turbulent transition regime, not a fully turbulent one. There are sery few heat transfer correlations developed for this regime, and this situation is further complicated by the fact that the flow is natural convection, in which the velocity / temperature profiles can be considerably distorted from those which would exist in turbulent forced convection. in addition, for Dittus-Boelter specifically--and all other similar heat transfer correlations (Colburn, McAdams, Seider-Tate)--the applicable range given is Re> 10,000 (and, it shoidd be noted, the aspect ratio range is L/D>60 for Ditrus-Boelter, which is also much larcer than exists in either the plant or the test article), so that the use of this correlation for the comparison / scaling of convective heat transfer appears to be incorrect. As a practical matter, however, this effect may be second order in many cases, and errors in its scaling may be of low importance overall. However, it is important that correlations with appropriate thermal-hydraulic and geometric ranges be employed in this study, because there may in fact be instances where the effects are not second-order. Address these concerns.
The following is additional information and clarification to Part e provided in a letter from the NRC dated March 24,1994.
W- WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION i
: 1. There appear to be errors in Revision 0 of WCAP-13963. " Scaling Imgie for the Core Makeup Tank Te st " ne text on page 2-12 stil! refers to use of the Ditrus-lioelter equation, though that equation is not employed in the discussion that follow s. There is also an inconsistency in the exponent on the Rayleigh (not Raleigh) number in Equations 2-37 and 2-30. and the term R in the denominator of Equation 2-37 is not defined (it appears to be a typographical error). Address these inconsistencies.
: 2. The staff questions the technical aspects of the approach used, both with regard to consistene) of definitions and appropriateness of the heat transfer correlation employ ed. The Rey noids number based on diameter is employed to demonstrate that both the plant core makeup tank (CMT) and the test article are in turbulent flow. While the relatively small diameter of the test article may make it look like a pipe, the use of a diameter- based Reynolds number for the plant CMT appears to be inappropriate, due to its very large diameter and extremely low flow rate; a length-based Reynolds number may be more appropriate, w hich also changes the criterion for laminar-to-turbulent transition.
Further, the staff concludes that both may not be in turbulent flow because (a) the value of Re abased on diameter) for the test article falls within w hat is generally considered to be a transition regime from laminar to turbulent flow, and (b) the very small aspect ratios of both the CMT (1.8/l) and the test article (about 6/1) would not allow the flow to become fully developed in any event. In light of the above reasoning, the use of length-based Grashof and Nusselt numbers (as in the correlation cited in Equation 2-37) appears to be appropriate, but it appears to be inco,nsistent with the use of diameter-based Reynolds numbers (including calculation of the Gr/Re ratio) to determine the single phase flow regimes. In addition, no information is included in the report to allow the staff to determine whether the correlation cited is appropriate for use over the range of thermal-h>draulic and geometric parameters characteristic of both the plant and test CMTs. The staff recommends that Westinghouse review this section of the report and reevaluate its approach to the question of wall-to-fluid heat transfer.
: d. The approach taken in the scaling analyses presented in Chapters 2 and 3 of the report mirrors that used by Oregon State University (OSU) in the scaling of the APEX facility. However, there do not seem to be any real conclusions draw n about the applicability of the test facility results to plant behavior, and, further, there is little or no discussion about what has been left out of the analyses at the start. For example, the scaling of both recirculatory and draining behavior is based on a one-dimensional momentum equation.
This is certainly applicable for the test facility, since multi-dimensional effects are suppressed by the small diameter of the test article. However, it may not be the case for the full-size CMT, where the diameter is large both in absolute terms and as compared to the tank length. De report does not address distortions due to suppression of multi-dimensional behavior in the test facility, nor is there an order-of-magnitude analysis showing the relative importance (or lack thereof) of multi-dirnensional effects. In addition, for many of those aspects of CMT operation not specifically analyzed, qualitative or intuitive arguments are used, with no supportingjustification. An example is found on page 32, in the discussion of liquid mixing and flashing effects during recircu!ation. The fact that things "seem like" they should behave similarly is not a substitute for a quantitative scaling study. Distortions and the uncertainties that they introduce into the scahng anal) sis should be dealt with in a quantitatise manner, w here possible. A/ dress these concerns.
440.52 2
[ W85tl!1gh0llSe

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1 NRC REQUEST FOR ADDITIONAL INFORMATION l
A
: e. The scaling analysis for draining behavior presented in Chapter 3 of the report does not provide an order-of magnitude analysis to separate the important phenomena / scaling parameters from the unimportant ones.
In aJdition, the scaling parameters derived for this mode of operation appear to vary over a very wide range (factors of 6 or more) during a given experiment, but the signi6cance of such a variation is not addressed. Also, the scaling of steam jet behavior should be updated to addrers the incorporation of the steam "distribut ar" at the CMT inlet. Dependence of the mixing length on steam diffuser design, scaling of the mixing length between the test facility and the Ch1T, behavior of steam when the diffuser is uncosered, and the means and impact of " scaling up" the diffuser should be addressed. The potential effects of non-condensible gases on condensation behavior and the resultant impact on Ch1T behavior should be addressed in a quantitative fashion. Furthermore, it appears that the mass conservation equation employed in Chapter 3 includes the assumption that all steam that enters the CMT condenses. Were this to be the case indefinitely, the CAIT would not drain, because some vapor (steam or non-condensible gas) will Gli the space vacated by the water as the tank empties. As the top layer of water in the CMT is heated, condensation will slow, and some steam (or other gas) will remain at the top of the tank, accumulating as the CMT drains. The equatiors used to assess draining behavior should explicitly reDect the physical processes that occur during that period of CMT operation; this includes not only condensation, but also the development, growth, and subsequent behavior of a thermally strati 6ed layer of liquid as draining continues, including the possible Hashing of the hot Guid as the system is depressurized. The proper scaling of these effects on CMT draining behasior should be addressed.
Proside a discussion of the signi6cance of the plots of predicted plant and model behavior, and dimensionless parameters and their ratios, that are presented in Chapter 3 of the report. It appears that for certain conditions at certain points in a given ty pe of experiment, the model will represent approximately the behavior expected in the plant, but it is not clear that such a conclusion can be extended to the range of conditions under w hich the CMTs are expected to operate, that the assumptions made (e.g ,
mixing depth) are realistic, nor that the idealized test conditions used for the analyses adequately represent the conditions that would exist during periods when CMT operation is most important. Address these concerns, and provide justi6 cation for these assumptions and for the selection of the test conditions.
: g. Discuss how the CMT test results will be related to CMT operation in other integral test facilities, and how the overall results will be implemented in the codes.

===Response===
: a. In the original SSAR AP600 design, the isolation valves on the CMT cold leg balance line and on tne CMT dhcharge lines were closed and would open on a CMT activation signal. This is what is stated on page 1-1 of the scaling report (Reference 440.52-1). Figure 1-1 of the scaling report also shows the sahes closed. In the AP6(X) design, presented to the NRC in April 1994, the pressurizer-to-CMT balance line was eliminated and the cold leg isolation valves are normally open. %ese changes have been documented in a design report to the NRC. Isolation of the CMT is achiesed by keeping the CMT discharge valves closed for normal operation. These salves open on a CMT actuation signal. These changes in the CMT design and logic will be included in a resision to the CMT scaling report. His revision will be issued by December,1994.
440.52-3 W
Westinghouse 1
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NRC REQUEST FOR ADDITIONAL INFORMATION I
: b. The nomenclature for the equations given in Section 2 is presented in Section 7-l. The term acis defined as the CN1T area and I is the heated length of the CMT.                                                               j l
: c. Response (including questions I and 2)                                                                             {
There is a t,spographical error in the exponent of Equation 2-39. The power should be 0.4 as given in Equation 2-37.
As suggested by the question, the heat transfer coefficients during the recirculation mode have been re-evaluated. A forced convection heat transfer coefficient was evaluated assuming flow over a flat plate with the hot CNIT liquid layer being the characteristic dimension. The correlation used is from Kreith and Bohn (Reference 440.52-2), a nd is given as i
Nu,= 0.332Re,'I Pr'13
                                                                                                                            )
for the local heat transfer and
                                                                                  'I Nut=0.664 Ret PR't3 i
l                                                                                                                           l
!                                                                                                                           I for the surface aserage heat transfer coefficient. Using the length or depth of the heated thermal laye,r as the characteristic dimension, the Reynolds and Grashoff numbers can be calculated and the ratio of Gr/Re can be compared for different thermal layer thicknesses. These calculations were performed for a range of CMT draining velocities w hich were calculated from the AP600 SSAR analysis as well as the draining velocities that are simulated in the CMT test. The SSAR CMT draining velocities are small, and range from 0.008 to,0.016 ft/sec, while the CMT test draining velocities will range up to a velocity of 0.127 ft/sec. The Gr/Re ratio w as calculated using the largest drain velocity of 0.127 ft/sec. The calculated ratios are given in a table below for different assumed thermal layer thicknesses. As the results indicate, the natural convection flow will dominate for all heated layer thicknesses such that the analysis presented on pages 2-11 to 2-13 of WCAP-13963 remains correct. 'lhese revised calculations using the flat plate forced convection heat transfer coefficient will
,        be included in a revision to WCAP-13963.
I 440.62-4 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION CAI.CULATING THE Gr/Re2 RATIO WAY THE TEST VELOCITIES Ileated Layer Thickness X                           Gr                     Re (.127 ft/sec)                 Gr/Re2 0.5                     3.49 x 1010                     44354.7                         17.7 1.0                    27.92 x 1010                     88713.94                       35.5 3.0                    753.9 x 1010                    266142.78                      106.34 Conclusion. Free convection dominates even for the fastest drain rates
: d. With regard to the conclusions drawn from the scaling study; there were conclusions peented at the end of the major sections of the report. For the recirculation phase, the Ch1T test captured all the phenomena of interest and would model the plant Ch1T behavior. For the draining portion of the transient, ther. were limitations that were indicated in the report. There is a time scale difference in the test Ch1T relative to the plant CN1T due to the wall thickness difference. This was specifically analyzed in the report. Also, the interfacial condensation correlation used tended to bias the results and introduce a scale effect. These points were also discussed in the report. The conclusion on the draining portion of the scaling analysis was that the CMT test will capture the phenomena of interest, for most transient time periods. There will be periods when the test will not accurately represent the plant CMT behavior due to the thinner walls used in the test.
However, as indicated in the report, there are sufficient time periods in which the tests is accurately representing the plant CMT such that this data can be used to develop or verify the models or correlations needed to model the AP600 plant CMT.
The question on the use of a onedimensional analysis for the CMT analysis and its validity is a separate issue.
The melusion of the Ch1T diffuser has reduced the occurrence of three dimensional effects for the majority of the Ch1T tank. Since the draining velocities are very small, the axial velocities in the tank are typically 0.008 to .127 ft/sec. Therefore, the flow down through the tank will be very one-dimensional. The region in which the three-dimensional flow can occur is at the CMT diffuser at the very top of the tank. The flow area of the dif fuser has been specifically sized to obtain a low liquid velocity (approximately 0.1 ft/sec) at the wall of the CMT both for the CMT test and the AP600 CMT. There is a recirculation flow pattern which is developed at the top of the CMT tank as either steam or liquid tiow radially enters the tank. The recirculation flow pattern has been observed to be confined to the top of the tank. It continues until the liquid at the tank iop is heated and the tank begins to drain. Once the CMT liquid is heated, steam condensation is reduced and the resulting steam velocity through the diffuser is reduced which reduces the mixing and recirculation. A more complete discussion of the recirculation and the scaling of the CMT diffuser will be included in the revision to the CMT scaling logic report WCAP-13963.
W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION There was a question on the lack of discussion on the flashing effects between the test CMT and the AP600 CM1 In the recirculation mode of operation, the development of the heated thermal layer is very similar between the test and the plant as seen in Figures 2-2 to 2-9 of the scaling report. Since the Huid conditions are presersed in the test relative to the plant CMT, if the depressurization rates of the test are similar to the plant the, flashing effects will be captured in the tests. The depressurization rate is a test parameter in the CMT tests and is controlled oser a range to approximate that in the plant. Herefore, the flashing elfects will be similar between the test and the plant. and this phenomena will be captured in the CMT tests.
: e. The CMT report will be revised to address the concerns expressed on the draining behavior or the CMT. The specific Questions raised in the question will be addressed and the CMT diffuser scaling analysis will also be included. The analysis that has been performed is quasi-steady in w hich a thermal layer thickness is specified l        and the mixing is assumed to be perfect within this layer. This is a reasonable approximation based on the 1
CMT dt.:a obtained to date.
l l
The flashing effects were not specifically developed for the CMT drain down analysis provided in Section 3 of the CMT scaling study. The mixing layer, which is what would flash once the CMT depressurized, was ranged over the thicknesses expected in the CMT tests and the plant. Since the fluid conditions are preserved in the tests relatise to the plant, the flashing effects observed in the test, over the range of the mixing layers, would be expected to be similar to the plant behavior. This will be discussed in more detad in the revised scaling report.
: f. The Figures presented in Section 3 of the report were sensitivity studies at conditions which were typical of                                                                                                         {
those conditions that the plant CMT would be expected to operate. The pressure selected is typical of the system pressure for a small break LOCA w ben the primary system pressure has equil;brated with the secondary                                                                                                          ,
system pressure. floth the liquid level and the mixing depths were parameters that were selected to examine I        the sensitivity of the test CMT to the plant CMT. As the plots indicate, the test will capture the thermal hydraulic phenomena of interest for selected time periods of the transient. This will be sufficient to develop or verify the models or correlations used in the plant for the CMT. In particular Figures 3-34 and 3-37 shows the dimensionless ir Froup ratios (test or model to plant) for wall condensation and liquid surface condensation. As figure 3-34 indicates, the model wall condensation is approximately the same as the plant for the first 700 seconds. After this time period the thinner walls of the model heat-up and the wall condensation significantly decreases.                                                                                                   Figure 3-37 shows the dimensionless ir ratios for the liquid surface condensation (model or test to plant). For very thin mixing depths, the ratio varies significantly as the water heats up faster in the plant calculation relative to the test calculation. This is caused by the choice of the condensation heat transfer correlation which is scale dependent as discussed in the report. However, for thicker                                                                                                      i more typical mixing depths, the agreement is much better and indicates that the test will simulate the surface                                                                                                        I condensation effects expected in the plant for a significant time period. The same behavior is observed in Figure 3-40 and 3-43 for a CMT tank level of 954.
Since design changes have occurred in the CMT balance lines and the initiation of the PRHR, The conditions presented in the existing scaling report will be resiewed to investigate if they should be supplemented by additional calculations. The additional calculations, if needed, will be included in the revised CMT scaling report.
440.52-G W
Westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION g.
The Ch1T teus are separate effects experiments u hich concentrate and isolate particular phenomena w hich will also occur in the integral systems tests such as SPES and OSU. The CN1T tests will be used to develop or verify existing models and correlations for wall and liquid surface condensation with the prescribed boundary conditions of the experiments. The Ch1T tests will be modeled with both NOTRUN1P and WCOBRA/ TRAC to examine the predictability of the existing models and correlations in these codes w hich are used in the SSAR analysis. If the agreement is inadequate, these models and correlations will be refined and improved to better predict the CNIT tests. Once satisfaction is obtained with the Ch1T data, the same model will be used to predict the OSU and SPES tests. In this fashion, the Ch1T tests are used for model deselopment while the integral tests are for model validation.

==References:==

440.52-1     WCAP-13963, " Scaling logic for the Core Atakeup Tank Test," transmitted to NRC via Westinghouse letter NTD-NRC-94-4068.
440.52-2      Kreith. F. and h1. S. Bohn Princirles of Heat Transfer, 4th Edition, pg 216-217.(1986).
SSAR Revision: NONE 1
440.52-7       1 W Westinghouse                                                                                                      1 1
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NRC REQUEST FOR ADDITIONAL INFORMATION mn   um
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:t:
Question 440.133 As a part of design features to mitigate air hinding of the RNS pump during mid-loop operation, Section 5.4.7.2.1 of the SSAR states that the RNS employs a step-nonle connection to the RCS hot leg. that will (a) substanti:dly lower the RCS hot leg level at which a vortex can occur in the RNS pump suction line due to the lower lluid velocity in the hot leg nonie, and (b) limit the maximum air entrainment into the pump suction, if a vortex should occur, to no greater than 5 percent. This value has been demonstrated experimentally. Provide a discussion of the actual design configuration of the AIWX) stepped nonle connection, the experiment (s) and associated data. as well as any analysis that demonstrate the adequacy of this design to minimite vortex fonnation and air entrainment into the RNS pump suction.

===Response===
Westinghouse test report. APWR-0452. "AP600 Vonex Mitigator Development Test for RCS Mid leop Operation" was provided to the NRC via Westinghouse letter NTD-NRC-94-4191, dated July 6.1994. This report describes a comprehensive test program perfonned to investigate the vortex behavior at the RNS line/ hot leg junction of the AP600 during mid-hop operation. The AP600 RNS line/ hot leg junction employs a step nonle design which confonns with the recommendations of the test report.
SSAR Revision: None PRA Revision: None i
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NRC REQUEST FOR ADDITIONAL INFORMATION
                                                                                                                      =     n:a 2       f Question 440.171 To address the concem of increasmg the potenti;d for operator error as a result of atided displays and controls in the control rmn of the high point vent systems. Section 5.4.12 of the SRP states that a human-tactor analysis should be pertonned taking into consideration the use of this infonnation by an operator during both nonnal and abnormal plant conditions, integration into emergency procedures and operator training, and other alanns during einergency and necil for priorituation of alanns. Confinn that the displays and controls of the high point vent will he mcluded as part of this human-factor analysis in the design of the displays and controls in the control rooln.
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Response:                                                                                                                                   l The design of the control room dnplays and controls for the high point vent systems will be included as part of the human-f actor analysis. Chapter 18 of the APNU SS AR describes the human tactor engineering design approach to be applied to the man-machine interf ace system design. Refer to sections 18.6.5. 18.6.7. 18.8.2.1.l.4 and 18.8.2.1.2 for descriptions ol' the human decision making model, the f unction based task analysis and their l
i apphcation to the design of the APNU man machine interf ace.                                                                                l l
l SSAR Revision: None
   .)
W- WestinEhouse J                                                                                                                             . . . _ . . . . . . _ _

NRC REQUEST FOR ADDITIONAL INFORMATION I
l     Question 440.245 l
l     Proside the following infonnation regarding the onsite de system:
: a.                 Section 8.3.2.1.1.1 of SSAR states that class IE divisions A and D have one battery bank each while disisions C and B have two. On the other hand, Figure C17-1 of the PRA shows that the 4 divisions have very similar structures. Explain the inconsistency.
: b.                 Tables 8.3.2-1 to 8.3.2-4 of the SSAR list the loads on the class IE buses. Are there similar tables for the non-class !E de buses?

===Response===
j     a.                The numbers of battery banks per class IE division as stated in SSAR Subsection 8.3.2.1.1.1 are accurate.
In the upcoming revision of the level 1 AP600 PRA, the de power distribution system modeline will reflect    l the current de power system design.                                                                          j l
: b.                The non-class IE de load assignments ar- shown on the drawings for the non-class IE de buses, SSAR Figure 8.3.2-3, Sheets I and 2.
SSAR Revision: NONE PRA Revision: The PRA will be revised by December 31,1994 l
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3 Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION A
Question 480.49                                                                                                      ,
l Note 12 of Sheet I of Figure 0.4.7-1 and Table 6.2.3-1 of the SS AR do not clearly describe the provisions for       j Type C testing for the metal-seated containment air tiltration system and chilled water system isolation vahes.      !
Provide a description of the design and Type C testing proposed for these sahes.        .
                                                                                                                         ]

===Response===
The containment air filtration system and chilled water system isolation valves subject valves are metal seated butterfly valves with a double seal around the periphery of the valves disk. Integral leakage connections to perform type C leakage testing are provided between the seals in order to pressurize between the seals and thus determine the valve leakage. The testing configuration is much like the leakage detection provisions provided for equipment hatches, fuel transfer blind flange or spare connections where double seals are provided with pressurization capabilities between the seals for leakage detection.
SSAR Revision: NONE W85tlngIl0LIS8

NRC REQUEST FOR ADDITIONAL INFORMATION Question 480.50 Generally, the P&lDs in the SS AR depict the test, sent, and drain (TV AD) v;dves provided for Type C testing.
Iloweser. the sersice air P&lD does not. 11 TVAD connections are not show n on a systein P&lD. does that inean that they will not h installed and that T.spe C testing of the isolation valves is not planned!

===Response===
}
The sersite air systern P&lD is presently under revision and will incorporate the necessary features to perhinn Appendix J. Type C testing. The features will be sitnitar to other typical penetrations with test vents outhoani of the outer containinent isolation valves. test connections inboard of the inner isolation vahes and a second test connection between the isolation vahes for penetrations that utilize a check valve as an inner containtnent isolation valve. SSAR Figure 9.3.1-I "Cornpressed and instruinent Air Systein Piping and Instannentation Diagr:un" will he revised anti included in Resision 2 of the SS AR.
l W
               - WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION gi: +g Question 480.52 Table 6.2.01 of the SSAR indicates that an x inch demineralized water transter penetranon is Type C tested with air in the lorward direction. Figure 9.2.41 does not depict the necessary TVAD connections. Also. Table 6.2.3-1 indicates isolation on a "T" signal: howeser. the valve is indicated to be a manual salve. Cl.irily these tables or the drawin g.

===Response===
The deminerated water system P&lD is presently under resision and will incorporate the necessary features to pertonn Appendit J. Type C testing. The features will be similar to other typical penetrations with test vents outbo.ud of the outer containment isolation salve. test connections inboard of the inner isolation check udve and a seconti test c onnection between the isolation valves. SSAR Figure 9.2.4-1 " Demineralized Water Transfer and Storage System Pipmg and Instrumentation Diagr.un" will be resised and included in Revision 2 ol' the SS AR.
The outboarti containment isolation valve is a locked closed manual valve and will not recieve a containment isolation signal.
SSAR Revision:
See the response to RAI 480.61 for revisions to Table 6.2.31 to reflect the abuse changes l
l 480.52-1 WM Westinghouse                                                                                                         l l
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NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 480.56 An S (P cnterion for use of reliel salves as containinent isolation barriers is that the serpoint be e 105 percent of the ontaininent design piewure. Continn that the reliet salves of Table 3.2-1 ol' the SS AR rueet this criterion. Will these salves open under sesere accident (Seruce Level C) conditions!
Response-The following rebet valses are identified as containtnent boundaries in SSAR Table 6.2.3-1:
CVS       Penetration P05                        Valve V056 CVS       Penetration P06                        Valve V042 RNS       Penetration P19                        Valve V021 SGS       Penetration P23/P24                    Valves V030A.B: V031 A.B: V032A.B The rehet salve set prewures for eath of the above valves will be in excess of both luW of containtnent prewure and a tontainnient pressure relating to ASME Service Level C stress linnts. The vah es will theref ore not open under desien basis accident conditions or severe accident design conditions.
SSAR Revision:
6.2.3.1.3      Additional Requirements M. Relief valves that serves as a pan of the containment boundary have a set pressure in excess of the 105 percent of contairuncnt design pressure.
PRA Revision: None l
480.56-1     l W=
Westinghouse l
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NRC REQUEST FOR ADDITIONAL INFORMATION EE HE E        @
Question 480.60 The terms 'm.mual" and " remote manual" are used in Table 6.2.3-1 of the SSAR to desenhe isolation device actuation modes. The statt understands that the tenn " manual." when used to descnbe the pnmary actuation rnode.
is used to mean manual operation from the control room. in addition. it is the staff's understanding that the term
    " remote manual." w hen used to desenbe the secondary mode of actuation. means manual actuation at a control station other than the rnam control nom. Confirm or clarify this understanding.

===Response===
The following clanficatmn has been added to the " Explanation of Heading and Acronyms for Table 6.2.31" Actuation Mode Pnmary/ Secondary:
Prunary closure mode of operation / Secondary closure mode of operation: Types:
manual:          manual manipulation at the valve (e.g. handwheel) self:            self controlled valve (e.g. check or relief valve) automatic:       power operated valve closes automatically on a safety related signal                      .
remote manual: power operated valve requiring remote operator action (e.g. from the MCR)
N/A:              isolation devices without manipulation capability (e.g. flange)
SSAR Revision:
A revised version of Table 6.2.3-1 is provided in response to Ital 480.61.
W85tiligh0llSO

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l NRC REQUEST FOR ADDITIONAL INFORMATION                                                                                        I 1
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Question 480 66 lius question pertains ni Westinghouse's staiement of confonnance to paragraph 6.2.1.1.A of the Standard Resiew               i Plan. "PWR Dry Containments. Insludmg Subannospherie Containments." that is identified on page 6-6 of Revision I to % CAP-13054. "AP6Ho comphance with SRP Acceptance Cnteria."
: a. What n Westmghouse's pmition relatise to retpiired margin between maximum calculated and design pr es su re .' 11 none is tonsidereil. then how does Westinghouse ensure there will be no ditterences between the cunent design stage and the "as built" of the actual plant.
: b. Proside justification for using best-estimate heat transler cocificients in Dil A calculations. Provide an unccHainty analysis for the worst case DIl A accident that takes these heat transfer coetticients into accouni.
: c. The stall understands that other parameters that are inputs to the DilA calculations are consers alise.
specific.dly boundary conditions and initial conditions such as input mass and energy release rates. Continn this unilerstanding.

===Response===
: a. The resuhs of the containment integrity an;dyses presented in Chapter 6 of the AP600 SS AR indicate 89 margin to the design pressure linut for the limiting case. The limiting cases base been reanaly/ed in reference 4XIL66-1 and indicate that there is approsunatelv 19M margin to the containment design pressure limit. These results indicate ample margin is available in the design. particularly m siew of the tact that the AP600 is a standaidized design under Part 52.
: b. The heat transler correlations used in the desien basis analysis calculations are taken trom open literature and applied sia the Westinghouse-GOTHIC code. The tree consection component of the heat transler correlation is calculated using the Mc Ad;uns correlation. The Colburn correlation is used for natural consection heat transler that n better representeil by a torced convection heat transler coetticient due to higher steam / air s elocities. These correlations are applied directly; however, there is the capability to apply heat transter coellicient multipliers to the calculation. To demonstrate the sensitivity to an uncertainty of 104 the limiting loss of coolant accident case was reanalyzed. The heat and mass transler coelficients were reduced to 904 of the calculated value and the transient was rerun. The reduttion in the heat transfer coetlicient should not allect the bbm dow n peak pressure, but should cause the containment pressure and teinperature to increase later in the transient.
The totitiinment pressure a:.d tetiiperature respinhe calculated with the reduced heat trainfer coetticients has been pioned and compared with the limiting loss of coolant accident case. The peak containment pressure iw his h occurs al the erld of blowdow n) incre;neil by ilt)$ psi a:Id the contatninent pressure at 24 hours increased by 0.6 psi tand is still less than halt the containment design pressure).
480.66-1 W
Westinghause

4 m NRC REQUEST FOR ADDITIONAL INFORMATION
: c. The inputs to Ilie Design ll. isis Analysis eak ul.itions, specifically the botandars and initial cornlitions, are conservat n e.
Nelef ens L's 4 xi tui- l . Westinghouse letter NTD.NRC.94 4174. "APNWI Panise Containment Cooling Sy stein Design liasis Aruly sis N1odels and Ntargin Awewinent". June 30. 1994 SSAR Revision: NONE 480.66-2 W
_ WestinEhouse

1 1
NRC REQUEST FOR ADDITIONAL INFORMATION
                                                                                                             #F     y W. !!M YVALUE                  1             0           0 SSAR Press (Ver 1 2) e         aYVALUE                    1             0           0 SSAR Press (907. HTX) 60 as W.
zum 50                               0                               ^
                                                  ,                             ^%;
              ~
m                                               ..
C      ~
Wm_
       ' 40
                                                                                           \
                                    '/r u
D       ~
30 lty  ~
: a.        _
20 h
                    ' ' ' ' ' ' ' '     ' ' ' ' ' ' "         ' ' ' ' ' ' ' '     ir iiiin         i i iiiin 10 0                      1                   2                    3                4                  5 10                   10                 10                     10                10                 10 Time (seconds) 480.66-3 W Westingh0use

NRC REQUEST FOR ADDITIONAL INFORMATION YVALUE                           1                                          0                         0 SSAR Temp (Ver 1 2) e                               a YVALUE                           1                                           0                        0 SSAR Temp (90% HTX) 300
                                                               ~
1 l
I                     v 250                                              l l                                                                                                                                                                                               s 1
                              ~
N                                                                                                                                        '%t%
i I
a
                             ~
200 w
l v                                _
l l                             E                                -
t                             o l                            r-150 l
                                                                                                                                                                                                                    \
l 100 o                     1                                                2                                         3                 4               5 l 10                                    10                                       10                                                  10              10                10  ,
Time                                                      (seconds)                                          1 480.66-4 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION En v
Question 480 68 This questnui pen.nns to Westinghouse's stateinent of confonnance to p.uagraph 6.2.1.2 ot the Stanitaril Resiew Plan.
        "Sobco:nparlinent Analpis." that is islentitietl on page 6-8 of Resision I to WCAP-13054. " AP(WHi Compliance with SRP Acceptance Criteria.'
        'Ibc innent design basis is beeil on the use of leak betore-break. Are the subcompartment analyses anti the anotiaieil wall capacities to be established by postulating the break of a 3" high energy line in each subcompartment.
icgarilless of whether that subcompartinent has any such lines in it!

===Response===
No. please see the response to R AI 210.76.
SS AR Res ision. NONii 480.68-1    1 W Westinghouse I

NRC REQUEST FOR ADDITIONAL INFORMATION
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Question 480 69 This question pert.uns to V 'esunchooseN statement of conh,rmance to paragraph 6.2.1.3 of the Standard Reuew Plan.
      "Mau and Energy Releas r' n.,iy. h- Postulated LOCAs." that is identitied on pages 6-X and 6-9 of Resision I to WCAP-13o54 " APN Wi ( omph.ma           ': SR P Acceptance Criteria."
: a. Proude the reference for the NRC-approsed TMD code Westinghouse intends to use for MA E analyses.
: h. For what reason is the TMD cale being used rather than a inore recognizable hcen ing code.'
: c. Document the specific assumptions used while performing this analysis,
: d. Identily the experimental data that will be used in this analysis.

===Response===
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: a. The TMD code is used to analy/e the pressures, temperatures. heat transler rates and mass flow rates as a tuottion of tune and location throughout containment. The TMD code is documented in WCAP-x077 tProprietary) and WCAP x07X (Non-propnetary). The TMD code received NRC approval IX December 1973 and i docmaented in a letter f rom D. B. Vawallo (NRC) to Romano Salvatori (Westinghouse).
: b. The TMD code is the NRC licensed code utili/ed by Westinghouse to perionn subcompartment analyses.
: c. The analysis specilie assumptions are presented in Table 6.2.1.2 and are discussed throughout section 6.2.1.2 in the APNW) SS AR
: d. No additional experimental data will be used in the subcompartment analyses of the APNHl. The test basis for the TMD code is provided in WCAP-5077.
WCAP-13054 wdi be unidilied to delete the re!'erence to TMD.
SS AR Resision: NONE W Westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION                                                                                                     !
f Question 492.5 Sections 4.3.1.6.2 and 4.4.6 of the SSAR prosides a brief description of the fixed in-core detectors.
: a.               Provide a more detailed description of the fixed in-core detector monitoring system on a functional and operational basis.
: b.               How does the AP600 thed in-core detecting system differ from the typical PWR movable detecting sy stem? Provide technical justification for this design.
: c.              Is the design of the proposed fixed in-core detecting sy stem approved by the NR(' If so, proside refer-ences.
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===Response===
,       a. The incore instrumentation system performs two basic functions:
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l            First, it prosides a means for monitoring, in an on-line real time mode, nuclear pow er distribution-related data from the reactor core. This data is used to provide a reliable inference of the actual three-dimensional nuclear power distribution in the core. Certain parameters, obtained by editing the resultant inferred nuclear power                          l
!            distribution, are then used both to calibrate the power range nuclear instrumentation for core protection purposes                    l l            and to proside guidance to the plant operators to ensure continuing compliance with the plant Technical Specifications and for enhanced fuel management.
Second, it provides in-vessel mechanical support for the core exit thermocouples used in post-accident core cooline monitoring.
l He signal output from each fixed incore detector is input to a signal processing device that cons erts the signals to digitized soltages and transfers them to a multiplexing device w here they are put onto a data highway.
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(            The incore instrmnent sy stem data processor receives the transmitted digitized fixed incere detector signals f rom                   l f            the signal processor and combines the measured data w ith analy tically-derived constants, and certain other plant l            instrumentation sensor signals, to generate a full three-dimensional indication of nuclear power distribution in j            the reactor core. It also edits the three-dimensional indication of power distribution to extract pertinent power I            distribution parameters outputs for use by the plant operators and engineers. The data processor also generates hardcopy representations of the detailed three-dimensional nuclear power indications.
: b. He AP600 fixed incore system differs from typical moving detector systems in that information pertaining to core power distribution is continuously monitored in the AP600 rather than the conventional method of monthly monitoring by means of incore flux mapping. With a movable detector system, incore power distributions ineasurements are taken once a month at a reference reactor operating condition. Between the time these measurements are performed, compliance to safety limits is assured by monitoring of a number of global 492.5-1 W
Westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION y
l parameters such as axial flux dif ference, or quadrant power tilt ratio. With the fixed incore sy stem, continuous monitoring of safety related core parameters such as I y and Fq is performed on line.
The in-core neutron detectors are fixed but removable and measure the neutron flux at representative locations throughout the core. The use of fixed detector assemblies capable of determining axial power profile eliminates the need for mosable in-core detectors and accompanying mechanisms. This simplifies plant equipment and reduces the maintenance effort.
: c.        The hardware components, specifically rhodium fixed incore detee. ors, utilized in this system are functionally simdar to components which are currently in use in operating plants. A Westinghouse designed incore instrumentation sy stem using fixed incore detectors is not yet approved. Information on the employment of fixed incore detectors in conjunction with an online power distribution monitoring system will be provided to the NRC to support the Final Safety Evaluation Report.
SSAR Revision: NONE 492.5-2 W-Westinghause

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NRC HEOUEST FOR ADDITIONAL If4FORMATIOt4 rs=:    .-
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Ouestion 952 82
       'l he I chinary 22. lona suhinitial on the AIYiw ilesign thanges naheates soine pilung shanges bir saf ety lines.
inslu'hne w heduhng thanges. Desenhe any t han,'es that will lv isnpleinented in the SPES 2 lacihty.

===Response===
All of the Alvul design thanges iniin the February 22. DN-i report w hit h isopacted the design and operation of the SPES-2 lacihty were unpleinented for the SPES 2 lacility starting wah inatrix test 500303.
SSAR Revision. None PRA Revision: None l
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